ML20245C755
| ML20245C755 | |
| Person / Time | |
|---|---|
| Issue date: | 04/20/1989 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| TASK-PII, TASK-SE SECY-89-123, NUDOCS 8904270373 | |
| Download: ML20245C755 (110) | |
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April 20, 1989 SECY-89-123 (Information)
Epor: The Commissioners 1 From: Victor Stello, Jr.
Executive Director for Operations Subiect: REVISED SEVERE ACCIDENT RESEARCH PROGRAM PLAN Purcose: The purpose of this paper is to inform the Commissioners of the staff's revised Severe Accident Research Program (SARP) which supports the tasks and objectives discussed in the staff's " Integration Plan for Closure of Severe Accident Issues," SECY-88-147. The principal objectives of this paper are:
- 1. To describe the major obje:tives and elements of the revised SARP.
- 2. To describe how the SARP activities relate to the Commission's policy, strategic goals, and other activities associated with closure of severe accident issues.
- 3. To describe how the SARP activities relate to those actions res11 ting irom the Commissicn's review of the FY 1989-1993 Draft Five Year Plan, namely:
- a. the staff's program for resolution of uncertainties, in the cource term, particularly as they impact closure of severe accident issues and ,
- b. the need for additional research on hydrogen transport and combustion beyond FY 1989 (provided as Appendix 1 to this paper).
1
Contact:
B. Sheron, RES, 492-3500 F. Costanzi, RES, 492-3525 F. Eltawila, RES, 492-3569
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Backaround:- For.the past 10 years, since the Three-Mile e Island accident, NRC has sponsored an-active
? research program'on' light water reactor. severe accidents as part of a multifaceted approach to reactor safety. In August 1985, the Commission issued a' Severe Accident ~ Policy Statement (50 FR 32138) in which the Commission concluded that, based on available information, existing
- plants posed no undue. risk to the public health and safety and.that there is no'present basis for'immediate action for any regulatory requirements for these' plants. However, based on NRC and industry experience with plant-o 6 specific probabilistic risk assessments'(PRAs),
the Commission was convinced of the need for both continuing research on severe accidents F and a systematic examination of each existing plant to identify any plant-specific vulnerabilities to severe accidents. These systematic examinations are now being accomplished under the Individual Plant Examination (IPE) program as described in the staff's Integration Plan for Closure of Severe Accident Issues (SECY-88-147) which was presented to the Commission in May 1988. That plan consists of six major elements:
- 1. Examination of existing plants for severe accident vulnerabilities (individual plant examinations).
- 2. Development of generic containment performance' improvements'with respect to severe accidents to be implemented if necessary for each of the-six containment types used in-the USA, namely, BWR Mark I, Mark II, and. Mark III; and PWR large dry, i subatmospheric, and ice condenser. !
4 3. Upgrading of staff and industry programs to improve plant operations.
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- 4. A severe accident research program.
- 5. A program to define how and to what extent vulnerabilities to severe accidents from external events need to be included in the -
severe' accident policy implementation.
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- 6. A program to ensure that licensees develop and implement severe accident management programs at their plants.
Discussion: The subject of this Commission paper is to describe the revised Severe Accident Research Program (SARP), Item 4 above, and how the.
revisions are designed to provide confirmatory information and technical support to the NRC staff in implementing the staff's Integration y
Plan for Closure of Severe Accident Issues as described in SECY-88-147. (The revised SARP The is provided as Attachment 1 to this paper.)
revised SARP addresses both the near-term research directed at providing a technical basis upon which decisions on important .
containment performance issues can be made, and the long-term research needed to confirm and refine our understanding of severe accidents.
In developing this plan, the staff recognized that the overall goal is to reduce the uncertainties in the source term sufficiently to enacle the staff to make regulatoryHowever, decisions on severe accident issues.
the staff also recognized that for some issues it may not be practical to attempt to reduce uncertainties-further, and some regulatory.
decisions or conclusions will have to be made with full awareness of existing uncertainties.
l l The near-term goals of the revised SARP, as l stated in Attachment 1, are:
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- 1. Provide the technological base for assessing containment performance over the range of risk-significant core melt events.
- 2. Develop the capability to evaluate the efficacy of generic containment performance improvements.
and the long-term goals are:
- 3. Provide an improved understanding of the range of phenomena exhibited by severe accidents that includes the impacts of l
I generic accident management schemes.
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- 4. Develop improved methods for assessing fission product behavior and availability -
for release in the event of containment failure at various phases of severe accident sequences.
These goals and their related issues set forth in the revised SARP plan are consistent with the Integration Plan for Closure of Severe Accident Issues (SECY-88-147). For example, in the near-term,'the revised SARP addresses- 1 L
issues and phenomenological uncertainties :
associated with accident sequences that lead to early containment failure: direct containment heating, BWR Mark I containment shell meltthrough, ex-vessel. molten fuel-coolant
' interaction in'BWR Mark II and III containments, and hydrogen combustion in BWR Mark III and PWR ice condenser containments.
For each of these issues, a program has been developed that attempts to obtain key data and-information needed to make an. informed decision on each issue within a fixed timeframe.
In Appendix 2 to this paper a summary is provided of how the results of the research to be conducted under the revised SARP will provide the required information needed to resolve near-term issues. The staff's current estimate is that the near-term programs will yield information that will provide tangible results within 3 years useable in the decicionmaking process in considering (a) the ,
likelihood of the phenomena in question, (b) various preventive or mitigative actions that can'either reduce or eliminate the probability of recurrence of the phenomena, or reduce or eliminate any adverse consequences associated with the phenomena, and (c) the likelihood that further research can or will substantially reduce uncertainties and the associated costs of such research.
DEVELOPMENT OF THE REVISED SARP During the development of the revised SARP, experts actively involved in severe accident .
research and severe accident phenomenology were drawn from laboratories, universities, industry, and NRC to consider the detailed technical issues and their status and to help the NRC staff define the needs for further research within the framework of SECY-88-147.
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SUMMARY
OF THE REVISED SARP-
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The ; plan' provides t a description of the general-approach to_the needed research. Naturally,
" .many of the details associated'with specific experiments or-analyses have not-yet:been
' developed and can only be developed after the.
specific research is identified. Therefore, ~
for'each.near term: issue, a detailed research plan will be developed, describing each t- experimental and analytical program and how
'these programs fit together.and' lead to'a
. resolution of the issue.
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In the remainder of this paper, the salient ~
features of the revised SARP-are brought to the Commission's attention.
- 1. Similitude (Scalina)
Because of the excessive costs associated a
with:large-scale experiments involving molten materials that cover the range of E potential severe reactor accident '
conditions of interest, most severe.
accident research11s' performed in small-scale facilities, many conducting only separate effects tests. Although conducting research in such a manner is not uncommon, questions about whether and to what extent the small-scale tests s reflect the fundamental realities of a reactor accident must be addressed.
Hence, a program to systematically investigate similitude (commonly called scaling)'has been started. The objective of this program is to develop methods and j criteria that will allow information from )
small-scale experiments to be confidently !
scaled up to full-size reactors. More ;
importantly, it will also tell us where information from small-scale facilities cannot be confidently scaled up. These methods and criteria woul'd be applied L
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before proposed new experiments were i funded. This program initiative is not intended in any manner to constrain closure of those other near-term issues defined herein (and in SECY 88-147).
Rather, it is a more demanding rigour being applied to similitude of the experimental work of smaller-scale which has prospective benefits to both the near-and long-term issues.
- 2. Late-Phase Core Melt Behavior The progression and consequences of severe accidents is highly dependent upon how and when the vessel lower head fails, and the amount, composition, and temperature of the molten core in the lower head at the time of failure. New research programs will be started that will be designed to allow us to get a much better under-standing of late-phase melt progression, ,
including lower head failure mode, within ;
the next several years. The result of this research will help confirm resolution of the direct containment heating and BWR Mark I containment shell meltthrough issues.
- 3. Code Development The current severe accident code development program involves the development of over a dozen codes to describe the various phenomena involved in, and phases of, a severe accident. For the most part, criteria have not been established regarding when development is finished. Under the revised SARP, care will be taken to ensure that criteria will be developed to assist the staff in deciding when a code is good enough.
Further, all codes are to be properly documented and the level of quality l assurance under which they were developed established. Future code development will be based on a structured approach such as the one currently being used for the thermal-hydraulic code development. This approach requires the developers to assess the current version of the code, identify potential deficiencies and proposed ,
improvements, and justify why it is l
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necessary to develop a new version of the code to incorporate the proposed improvements (e.g., how is our perception of safety or risk expected to change, what is the likelihood the proposed improve-ments will in fact correct the deficiency). Employing this approach to code development along with criteria to help determine when the codes are " good enough," is expected,to ensure a balanced approach to uncertainty reduction that will prevent the development of any one code to the point where its uncertainty has been reduced far below the uncertain-ties associated with the phenomena dominating the risk.
Finally, there are currently four in-vessel core melt progression codes under development.
These are RELAP-SCDAP (INEL) , TRAC-MELPROG (SANDIA), BWRSAR (ORNL) , and MELCOR (SANDIA).
The need to continue development of all of these codes is not clear. Each code will be critically examined, its capabilities and other attributes determined, and a decision will be made as to which codes should continue to be supported.
- 4. Lona Term Pesearch Simultaneous with the near-tern effort, the staff will continue to pursue a longer-term, '
broad-based research program aimed at reducing uncertainties in estimates of severe accident risk. Although resolution of issues associated i i
with this broader-based rescerch are not considered of immediate consequence as are the issues associated with the near-term goals of the SARP, nevertheless they are not insignificant contributors to risk uncertainty.
In pursuing this long-term research, the staff will apply many of the same elements of the near-term program. For example, code development will follow the approach previounly outlined, and scaling methods and criteria will ,'
be applied to ensure fidelity ofInsimilitude addition, in the experimental undertakings.
the long-term programs will be periodically reviewed to ensure that their continuation remains essential to understanding severe accidents and reducing severe accident risk uncertainties. The operative criterion will be
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9 whether there is a high likelihood that their continuation will result in a significant reduction in risk uncertainty.
Accordingly, for the long-term program some new programs are likely to be initiated and some existing _ones will likely be either terminated or revised.
Resource The NRC resources associated with carrying out Commitment: the revised SARP over the next three years including regulatory closure of those severe accident issues (of immediate priority that need to be addressed in order to achieve the near-term goals for regulatory closure as set forth in SECY-88-147) as summarized in this paper are consistent with the budget set forth in the current Five Year Plan approved by the Commission.. This is approximately 10 FTE of staff effort and about S18M of contractor assistance in FY 1990, 10 FTE of staff effort and about $23.4M of contractor assistance for FY 1991 and 10 FTE of staff effort and about
$23.8 of contractor assistance for FY 1992.
Resource needs beyond FY 1992 for the long-term program will be submitted for Commission approval as part of the normal budgetary process. It should also be noted that some of the tasks identified in the attached revised SARP are being carried out under the accident management research program element and are presented in the revised SARF to provide a comprehensive description of the integrated program. There are no duplicative efforts in any of these elements.
Increases in funding levels to carry out the revised program are not proposed in the foreseeable future. In revising the program, in addition to identifying new or reoriented
.. prograns, the staff plans to either terminate or reduce funding of existing programs consistent with the revised program emphasis.
With regard to advanced LWRs, the staff is not aware of any features of the evolutionary LWRs under development that are unique with respect to severe accident analysis and require special research programs. As such, no funding has been allocated for work in this area. However, the staff will keep abreast with the development of the evolutionary LWRs and notify the Commission if the current perception
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1 10-e changes. The' staff intends to advise the -
commission periodically on the SARP. progress toward achieving closure'of the near-term severe' accident issues and as appropriate, the status of the long-term program..
Schedulina: This paper is scheduled to be considered at an open meeting on May 2, 1989.
C V ctor-Ste o, .
Executive Director for Operations
Enclosures:
L ~ Appendices 1 and 2 Attachment 1 provides Revised Severe Accident
- Research Program DISTRIBUTION:
Commissioners OGC OIA GPA
-REGIONAL OFFICES
-EDO ACRS-ACNW t
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Appendix 1 )
Hydrocen Transport and Combustion In a memorandum dated September 13, 1988, from Victor Stello, Jr.,
to Eric S. Beckjord, " Actions Resulting from Review of the FY 1989-1993 Draft Five Year' Plan," the staff was requested to provide EDO with a Commission Paper that justifies work on hydrogen transport and combustion beyond FY 1989. The memorandum also requested the staff to identify the cost of the proposed work. In a memorandum dated November 8, 1988, from Eric S. Beckjord to Victor Stello, Jr., " Commission Papers on Source Term Uncertainty Resolutions and Hydrogen Transport and Combustion Work," the staff indicated that a comprehensive SARP will address both the source term uncertainty resolution and hydrogen transport and combustion research.
Attachment 1 to the Commission Paper transmits the revised SARP.
The purpose of this appendix is to summarize the hydrogen transport and combustion issues.
Hydrocen Behavior for Dearaded Core Accidents The major concerns regarding hydrogen in LWRs are that the static or dynamic pressure loads from hydrogen combustion and detonation may breach containment integrity, or that safety-related equipment may be damaged as a result of either pressure loads or high temperatures. To assess the possible threat to containment and safety-related equipment, it is necessary to understand how hydrogen is transported and mixed within containment and to determine the likelihood of various modes of combustion. The hydrogen behavior issues have been extensively investigated since 1979, but several important areas of uncertainty remEin:
(1) high-temperature /high steam concentration combustion and (2) deflagration-to-detonation transition Research programs on these issues of combustion have been sponsored in the U.S. by the NRC and the nuclear industry as well as in the international community. However, the combustion processes are i
sufficiently complex such that many aspects are still not well understood. The resulting uncertainties in the threat to
- containment integrity are unlikely to be reduced significantly by the existing research programs. The intent of the revised SARP It should be (attachment noted that in 1) is tocases, some reduce these uncertainties.
reduced uncertainties are not required to make a near-term regulatory decision. Each category of uncertainty is discussed below.
Hich-Temperature Combustion The Zeldorich-von Neumann-Doering (ZND) chemical kinetics theoretical model developed under NRC sponsorship predicts that Al-1 1
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the detonation of off-stoichiometric hydrogen-air mixtures and on all hydrogen-air-steam mixtures. It is believed that the steamThere inerting effect is reduced greatly at elevated temperatures.
is a limited supporting data base for this phenomenon. Experiments are necessary to resolve the uncertainties associated with high temperatures and steam concentrations typical of those likely to be encountered in severe accident scenarios. It is also necessary to extend or develop more mechanistic models to predict, by either extrapolation or interpolation, the temperature sensitivity of hydrogen-air-steam mixtures t!at have not been tested.
The small-break LOCA and TMLB scenarios are two examples of high-temperature hydrogen-air-steam mixtures that may exist below the auto-ignition temperature (550 C minimum value for stoichiometric, hydrogen-air mixtures). There are two aspects of the high-The first temperature /high steam concentration combustion problem.
is the injection of high-temperature hydrogen and steam mixtures that auto-ignite upon contact with pre-existing and premixed hydrogen-air-steam mixtures. The second aspect is the injection of hydrogen and steam mixtures at elevated temperatures that do not auto-ignite upon contact with pre-existing hydrogen-air-steam mixtures. This allows the possibility of a premixed condition to form and a subsequent deflagration or detonation. In both cases, the competition between chemical reaction rates and physical mixing rates will ultimately determine the ensuing combustion mode. Data are needed to determine the hydrogen-air-steam flammability limits and steam inerting criterion at elevated temperatures if reliable predictions are to be made as to the likelihood and the potential threat resulting from various combustion mode (s) for a wide range of accident conditions.
Therefore, the research approach is to determine hydrogen-air-steam flammability limits, volumetric oxidation (chemical reaction) rates, physical mixing rates, and the competition of these rates during the injection of high-temperature hydrogen and steam mixtures into cooler pre-existing and premixed hydrogen-air-steam mixtures.
A draft proposal is currently under consideration by the NRC that addresses feasibility; construction design, cost, and schedule; and an experimental test plan. It is our current estimate that if a facility is needed, it can be constructed by early FY 1990 and that experimental results can be generated by late FY 1991 or early FY 1992. Final data reduction and formal documentation should be available in late FY 1992.
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4 Containment Loads for Detonations Deflagration-to-Detonation Transition (DDT) l Direct initiation of a detonation would require a concentrated i j high-energy source for insensitive, steam-diluted mixtures. This is not considered a credible mechanism of initiation by almost all researchers. However, it is possible to initiate a flame with a low-energy source such as a spark or a glowplug, and the subsequent i propagation through orifices and around obstacles such as pipes can result in flame acceleration that culminates in a transition to detonation.
The possibility of DDT in realistic and prototypic containment geometries and conditions needs to be resolved. The uncertainty in this area has increased because of recent experimental and theoretical results that indicate an increased likelihood of 1 detonations at high temperatures and large steam fractions as !
discussed earlier. The current data base on flame acceleration and l DDT suggests that the mixture composition, obstacles, and venting are all important factors. These uncertainties result from a lack of experimental data on flame acceleration and DDT for conditions that include the effects of steam dilution, elevated temperature, ,
large-scale and prototypical obstacle types, and spacing.
At present, no reliable model exists to predict or extrapolate DDT results from small-scale experiments to containment scale. Reactor safety studies and fundamental combustion research is being carried out in the Federal Republic of Germany and Canada. These data along with data generated for space shuttle application will be applied to reducing the uncertainty associated with DDT as well as in assessing the potential threat of DDT to containment integrity.
Specifically, this data base will then serve as the basis to improve or develop correlations and models to allow extrapolation of experimental results to reactor scale and acc3 dent conditions.
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Appendix 2 Elements of the Near-Term SARP Related To Containment Challences As stated in the Commission paper, a program has been developed that attempts to obtain key data and information needed to make an informed decision on each issue. A more' detailed description of this program is presented in the revised SARP (Attachment 1) and is I summarized below.
(1) Depressurization and Direct Containment Heatina The conduct of the research enumerated under Issue 2, "Depressuri-zation and Direct Containment Heating," would provide the technological base for assessing the primary containment integrity to the threats posed by DCH and in assessing mitigating strategies.
In particular, executing the DCH research will enable the staff to gather the information needed to make an informed decision regarding the likelihood of a DCH threat to containment integrity.
If such a threat exists, the staff will also be able to make recommendations on the feasibility and the effectiveness, including any adverse consequences, of depressurization on the reduction of the risk associated with DCH. The assessment would always include the current state of operator training and the dominant accident sequences anticipated. In addition, the proposed research will examine the effectiveness of water on quenching the melt and subcompartment effect on reducing the consequence of DCH.
(2) BWR Mark I Containment Shell Meltthrouch It is widely agreed that adding water under severe accident conditions is a desirable strategy to attempt to quench the debris and keep it coolable and hence may prevent Mark I containment shell meltthrough. Moreover, water can act as a scrubbing mechanism for l fission products and could substantially reduce the radionuclides l released even if containment shell meltthrough were to occur.
l Executing the research activities described in the revised SARP l will enable the staff to determine the mode and timing of the RPV l lower head failure and hence identify the various melt conditions, l , i.e., quantity, composition, and temperature. The staff will also
! be able to better quantify the effect of water on the drywell floor on melt spreading and on the heat transfer between the melt and the containment shell.
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t (3) BWR Mark III and PWR Ice _ Condenser Containment Hydrocen Combustion The risk-significant potential of early containment failure by L hydrogen burns and/or hydrogen detonations will be assessed using data.and models provided by the SARP, namely, the HECTR and HMS-BURN,-codes which have been validated. Relative to hydrogen detonation issues, the recently developed model ZND.will be used for a. prediction of the detonability of a given mixture of hydrogen, air, carbon oxides, and steam.
(4) BWR Mark II and Mark'III Fuel-Coolant Interactions A potential mode of containment failure involving the interaction of the molten core with water may be important for those unique Mark II and Mark III designs in which there exists direct pathways between the area underneath the vessel and the water from the suppression pool.
The research to be conducted will enable the staff to address some well-defined questions dealing with key parts of ex-vessel molten core-water interactions and the potential that such interactions could lead to failure of the containment at a major penetration.
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I REVISED SEVERE ACCIDENT RESEARCH PROGRAM PLAN 1
i j U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research I
April 1989
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l SEVERE ACCIDENT RESEARCH PROGRAM PLAN Revision 1 i r
APRIL 1989 I. Introduction i II . - Goals l III. Meeting Near-Term Goals !
l Issue 1 Scaling j Issue 2 Depressurization and DCH l Issue 3 BWR Mark I Containment Shell Meltthrough Issue 4 Adding Water to a Degraded Core ;
Issue 5 Use and Status of Severe Accident Models (Codes) !
IV. Meeting Long-term Goals !
Issue L1 Modeling Severe Accidents Issue L2 In-Vessel core melt progression and hydrogen generation l Issue L3 Hydrogen Transport and Combustion 1 Issue L4 Fuel-Coolant Interactions Issue L5 Molten Core-Concrete Interaction (MCCI)
Issue L6 Fission Product Behavior and Transport Issue L7 Fundamental Data Needs ,
TABLE 1 SARP Milestones and Estimated Costs (3 year projection)
Appendix A Background on SARP and Relationship to other l
Elements of SECY-88-147
, Appendix B A Severe Accident Scaling Methodology (SASM) l
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i Severe Accident Research Procram Plan
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I. INTRODUCTION In a memorandum dated July 8, 1988, the Director of the Office of Nuclear Regulatory Research charged the Division of Reactor and Plant Systems (now the Division of Systems Research) to develop "a detailed plan identifying individual goals, discrete products, and anticipated schedules for the Severe Accident Research Program (SARP) that clearly demonstrates the implementation of the SARP portion of the staff's Integration Plan for closure of Severe Accident Issues (Integration Plan) (SECY-88-147). [ Preparing such a plan will require) a detailed review of the current activity of the SARP against the framework of the Integration Plan and the detailed activities and schedules of the other five elements of the Integration Plan, [together with) an assessment of what redirecting /reprioritizing/ rescheduling within the SARP may be needed to fulfill its portion of the Integration Plan and how and when such can be accomplished." This revised Severe Accident Research Program plan reflects the evolution of the severe accident research program and was developed in response to that July 8 memorandum. It will be used to guide the formulation and conduct of severe accident research, addressing in the near-term those issues pertinent to the implementation of the Integration Plan:
viz., Individual Plant Examination (IPE) and Containment Performance Improvement (CPI) activities. The plan also will be I-1
used'to guide the formulation and conduct of the program of long-term research needed to support the NRC's accident management activities and to confirm the Commission regulatory decision on severe accident issues. The plan provides a description of the general approach to the needed research. Naturally, many of the details associated with specific experiments or analyses have not yet been developed and can only be developed after the specific research is identified. Therefore, for each near-term issue, a detailed research plan will be developed, describing each i
experimental and analytical program and how these programs fit together and lead to a resolution of the issue. The plan may also give an impression that we are just now entering the arena of severe accident research; the facts are quite the contrary. A tremendous amount of severe accident research now exists and this plan is intended as a needed step to the critical and focused re-examination of our further needs and the regulatory questions involved.
I This plan is organized as follows.Section II presents the goals In Section III, the of the SARP and a sketch of its structure.
work directed at achievement of the near-term goals articulated in Section II is described. Research dealing with the longer-term goals is covered in Section IV. Appendix A to this plan provides background information on the existing SARP and the relationship of SARP to other elements of SECY-88-147, and Appendix B is an I-2
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II. GOALS In developing the revised Severe Accident Research Program (SARP) plan, the staff recognized that the overall goal is to reduce the uncertainties in source term sufficiently to enable the staff to make regulatory decisions on severe accident issues. However,.the staff also recognized that for some issues it may not be practical to attempt to reduce uncertainties further, and regulatory decisions or conclusions will have to be made with full awareness of existing uncertainties.
As the Severe Accident Research Program represented in this plan is a goals-oriented program, it is critical that such goals be clearly stated at the outset. The staff expects to use safety goal policy and objectives, as appropriate, in determining what potential improvements in the technological base are needed for clocure of severe accidents and in order to ensure that the safety of existing plants are reasonably consistent with the safety goals . The goals l
l of the revised SARP are as follows:
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- 1. Provide the technological base for assessing containment performance over the renge of risk-significant core melt events.
- 2. Develop the capability to evaluate the efficacy of generic containment performance improvements.
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. 3. Provide an understanding of the range of shenomena' exhibited ,
by severe accidents that includes the impacts of generic accident management schemes.
- 4. Develop improved methods for assessing fission product behavior.and availability for release in the event of containment failure at various phases of severe accident sequences.
Goals'"1" and "2" above are primarily near-term, while goals "3" and "4" are longer-term objectives that aim for additional depth of understanding of both accident evolution and final consequences.
These goals =are consistent with NRC's Integration Plan for. Closure.
of' Severe Accident Issues (SECY- 88-147), and the guidance provided
- in Generic Letter 88-20 for Individual Plant Examinations. It should be noted that there are not two SARPs, a near-term and a long-term. Rather the revised SARP draws from and focuses specific tasks of the continuing program of severe accident research experiments and analyses to address the near-term implementation of SECY-88-147.
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A. Near-Term Work Subsequent to the Three Mile Island Unit 2 accident in 1979, the U.S. Nuclear Regulatory Commission undertook a broadly based I l
research program to develop an understanding of severe accident behavior. Over the past decade, major experimental and model (code) development programs have been performed that provide a greatly improved understanding of severe accident phenomena, reflected in an ability to model those phenomena. Now, as the commission is preparing to close on severe accident regulatory questions, the immediate priority of the NRC's SARP is to support the closure process. That process' is displayed schematically in Figure 1. Hence over the next 3 years a significant portion of the severe accident research effort will be directed to issues that relate to the dominant topic area of the Integration Plan, viz.,
the accident sequences that lead to early containment failure:
direct containment heating (DCH), BWR Mark I containment shell meltthrough, molten fuel-coolant interactions in BWR Mark II and Mark III containments, and hydrogen detonation in BWR Mark III and In SECY-88-147, the staff identified the steps that each licensee is expected to take to achieve closure on severe accident for its plant, namely:
1- Completion of the individual plant examination (IPE) and identification of potential improvements 2- Development and implementation of a framework for an accident management program that can accommodate new information as it is developed, and 3- Implementation of any Generic Containment Performance Improvements with respect to severe accidents.
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issues for accidents in which the core is degraded but is not into .
i the severe accident domain have been extensively investigated which i
resulted in the. resolution of USI A-48 " Hydrogen Control Measures-and Effects of Hydrogen Burns on Safety Equipment". The additional hydrogen research discussed'in this revised SARP is intended to address the hydrogen threat from severely damaged cores and-will serve to reduce uncertainties in the threat to containment for a integrity resulting from various hydrogen combustion mode (s) f~
wide range of accident conditions. In the near-term, severe accident research will also support the resolution of broader, more general questions regarding the response of various containment Task descriptions have been designs to severe accidents.
developed for the near-term work on these issues and are presented in Section.III of this plan.
The majority of the near-tern severe accident issues are considered relevant to evolutionary LWRs currently under review by the NRC. Therefore, these near-term efforts are equally important to evolutionary LWRs. We are currently not aware of any features of the evolutionary LWRs under development that are unique with respect to severe accidents and As such, no funding has been require special research programs.
However, we will keep abreast allocated for work in this area.
with the development of these evolutionary LWRs and notify the Commission if our current perception changes.
II-4
B. Lona-Term Work Achievement of the longer-term goals demands a broadly based severe J accident research program. To be responsive to regulatory needs, particularly confirmation of closure of severe accident issues, that program must both explore severe accident phenomena and develop the methodologies appropriate to-quantitative assessment of severe accident risks. The plan to guide'such a program must reflect a balance among the risk importance of phenomena being investigated, the desire for technical completeness, the expected likelihood that the research will result in a significant reduction in the uncertainty of risk, and the need to promote the maintenance of technical expertise and analytical capabilities. The features of the SARP that address the long-term goals are presented in Section IV of this plan. Figure 2 depicts the elements of this revised SARP and the relationships among them.
C. Results of Previous Work Review of information available from previous and ongoing research including that sponsored by U.S. industry or foreign organizations is being undertaken by the staff and expected to be completed by
[ end of FY1989. The purpose of this review is to identify and redirect the.research program as appropriate to ensure that the needed information to reach closure or resolution of severe accident issues important for regulatory decisions is developed.
i II-5 l
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af, s This review.will.also ensure'that'the various research projects are consistent and well integrated among themselves, have a common goal
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' that-the long-term confirmatory research is properly focused.
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.J III. MEETING NEAR-TERM GOALS This section of the SARp plan presents those issues of immediate priority that need to be and can be addressed in the next few years in order to achieve the near-term goals set forth in Section II.
These are the issues that are pertinent to the NRC's programs of individual plant examinations and containment performance improvements (direct containment heating (DCH) and hydrogen
~
transport and combustion, issue 2, and BWR Mark I containment shell meltthrough, issue 3). The research to be performed to address those issues is also presented. This research is primarily directed to assessing containment performance under severe accident conditions and is constructed to enable as definitive a judgment as is possible to be made about the potential threats and likelihood of early containment failure in the event of a severe accident given the present paucity of (or sparse) data describing core melt progression. Also included in this section is a discussion of the issue of scaling of severe accident experiments and analyses (issue 1). Obviously, the question of verisimilitude in severe accident research is critical, and scaling is one element of the same.
Whether the correct physical and chemical processes are being investigated is another and is appropriately dealt with on a case-
.by-case basis. But scaling as an issue is common to all areas of severe accident research since full-scale experiments simply are not feasible. For this reason, a systematic examination of scaling of severe accident experiments and analyses is included among the key elements of the near-term research. This element is III-1
,* ,. e
- not intsndsd to constrain clocuro expncted of the other naar-term issues herein but is expected to be of benefit to decisions made on recommending further experiments over the near- and long-term
. periods addressed in this plan.
In addition, this section contains a discussion of the issue of molten fuel-coolant interactions (issue 4) and a discussion of the
. issue of the state and use of severe accident codes (issue 5). The molten fuel-coolant interaction issue is included here because it is directly relevant to considerations of intentional reactor i vessel-depressurization to cope with DCH, as well as low reactor coolant system pressure accident sequences in general.
Specifically, the question of what happens when water is added to molten fuel (in-vessel or ex-vessel) appears to be significant to understanding accident management and some severe accident sequences. How to answer that question is not immediately obvious, and some initial work needs to be done to allow-development of a rational approach to the research. As for the issue of codes, it is recognized that code development is integral to the achievement of long-term goals, since the understanding of severe accident phenomenology ultimately is reflected in those calculational tools.
However, achieving the near-term goals will necessitate, reasoned judgments based essentially upon our present understanding of severe accidents, and an assessment of just how well present codes describes the key behavior of severe accidents. The confidence j l
that can be plact.' '4n the severe accident codes also directly j
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III-2 l
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a o, itranslates into: confidence that can be vested in the near-term' J l
For this-reason, the code a judgments about containment performance. .\
- assessment issue is included in this.section.
Severe' Accident Phenomena - Core Melt Accidents Before the-issues and research tasks to address'them are discussed,
~
a general description of the-important phenomena associated with core melt accidents to be dea'lt with in the near-term is presented.
This is done to provide a perspective'and context for the issues
'and tasks.
In' light water reactors, core'nelt accidents can occur either at
~ ~1ow reactor coolant system pressure or at high reactor coolant system pressure. At low pressures, the core heats up more or less independently from the rest of the primary system, and the process is accelerated by metal-water reactions. At high pressures, energy redistribution by steam from natural circulation may be significant, transferring core heat to the remainder of the system, including upper vessel internals and, for PWRs, steam generators.
This transfer of energy affects the progress of the accident in L
/ several ways. It delays somewhat the onset of gross core degradation. It affects the steam availability in the metal-water l-
' reaction process, tending to limit hydrogen generation. It affects radionuclides release and transport within the primary system.
Perhaps most importantly, it also raises the possibility of III-3
competing high temperature and stress related of failures of the primary system pressure boundary (e.g., RCS piping, steam generator tubes) in PWRs with attendant depressurization, hence avoiding the possible ejection of molten core material from the reactor vessel while under high pressure. Current understanding suggests that for BWRs natural circulation is restricted by the fuel assembly wall channel boxes. Although natural circulation may affect the core melt progression, failure of the reactor coolant system due to excessive heating from high-temperature naturally circulating gases would not be likely. However, because of the enhanced reliability of the Automatic Depressurization System (ADS) suggested from the containment performance improvement program, high-pressure scenarios are not likely to be dominant contributors to early containment failure in risk assessment of BWRs. Hence, high-pressure scenarios in BWRs are not being addressed at this time.
Although the failure of the reactor coolant system (RCS) by natural circulation appears to be possible, the ability to predict its occurrence has not been established to the point that confident conclusions regarding the likelihood of RCS failure can.be made.
Moreover, while analyses to determine the extent to which natural circulation occurred in the TMI-2 accident are under'way, it must l
be noted, however, that no evidence of high temperatures in the l
range of interest to cause failures in the primary system components has yet been seen. Thus, competing RCS pressure boundary l
l III-4 l
1 _ . _ _ _ _ - _ _ _ - _ -
., ; o fail'ures from high temperature cannot be. presumed with a high
. degree of confidence, and scenarios involving. gross core degradation or core melt at high or intermediate RCS pressure must be. considered. Since such high pressure'PWR scenarios have the potential to produce ejections of molten core material.into the lower' containment following lower vessel head failure, a variety of concerns'regarding,the possibility of containment
- overpressurization have been raised, in particular, direct-containment heating. In fact', the high-pressure scenario seems to
' influence risk in terms of the variety of mechanisms with potential to overpressurize the~ containment and thus importantly contributes to the uncertainties-in predicted risk. . In addition, core melt accidents at low RCS~ pressures that lead to early containment failure have been identified as potentially important to risk, in particular, BWR Mark I containment shell meltthrough.
The following observations are intended to supplement the above
' general perspectives:
V
- 1. At this time, competing primary boundary failure locations, size, and timing are highly uncertain, particularly considering the role of relocated fission products, and the possibility of loss of integrity of steam generator tubes.
l III-5
+ ,
Core slump at high-pressures may yield significant b
2.
fuel-coolant mixing. Triggering of massive steam explosions e at such high pressures is generally considered highly unlikely by the experts, but the potential driving forces associated with such core slump interactions may not be altogether ignorable in the process of understanding fission product relocation and the potential for altering the accident progression.
- 3. Natural circulation may for example, cause a weakening of the upper internals and upper vessel head (e.g., seals), with <
I concomitant result that competing relief paths may be formed that influence both the fission product relocation and the potential for direct containment heating.
4.
Lower-vessel head specific failure modes and the timing remain highly uncertain. Local failures of instrument guide tubes could occur. However, melt attack of the vessel lower head forging (in the absence of local penetrations) could lead either to a local failure of the forging or to creep rupture of a large portion of the lower vessel head when heated to 600
.- to 700C. Large failures at high reactor coolant system pressure can potentially create thrust loads and threaten failure for example,through overstress or tearing at piping !
junctures.
III-6
.' 1, 5.- At high pressure, a large coherent failure of the lower head would likely be accompanied by prompt expulsion of melt, steam, and hydrogen. High-pressure expulsion has been postulated to lead to direct heating'of the containment (DCH) atmosphere and an increased potential for hydrogen combustion and perhaps even detonation. -
6.- The ex-vessel course of accidents in which reactor vessel !
breach occurs with low pressure in the reactor coolant system is highly dependent on the mass, thermal and chemical properties of the melt upon exit from the reactor pressure vessel. Direct attack of the Mark I containment shell; molten fuel-coolant interactions in the BWR Mark II and Mark III containments and hydrogen combustion in BWR Mark III containments are examples of where low system pressure scenarios are likely to dominate the probability of a severe accident but also give rise to early containment failure threats.
Issue 1 - Scalina Full-size severe accident experiments are impractical or simply not l
feasible, for many reasons, particularly costs. Hence severe l
accident research must rely on small-scale experiments that are I carefully crafted so that the conclusions drawn from those experiments about the phenomenology of severe accidents can be appropriately applied to predict severe accident progression in typical nuclear power reactors. The confidence in this application III-7 l.
._ -__ _ _ _ _ _ _ _ _ _ _ _ _ ._- __ _ __ - ___ _ _ __ _-_- _ _ _ _____ - ____ - _ - -_____ -- - ______ _ a
4 .
is greatly enhanced through scaling analysos. A key element of ths
-SARP,~therefore, is a scaling methodology for severe accident experiments. In those cases where an acceptable methodology can be .
developed, it will be applied to relevant experimental programs.
As necessary, recommendations regarding revisions to experimental programs will be made and needs for new facilities will'be addressed in terms of the regulatory value in pursuing this research. A useful scaling methodology should be capable of providing the following:
- 1. A scaling rationale and similarity criteria.
- 2. A procedure for conducting comprehensive review of facility ,
design, test specifications, and results.
3.- A measure or index to indicate the applicability of correlations or models based on test data from sub-scale l
facilities to full-scale nuclear power plant conditions.
- 4. Quantification of the effects of scale distortion.
- 5. Quantification of the effects associated with extrapolating l
correlations or models beyond their data base. Thic l
information is needed for quantifying code uncertainty, III-8
o .
- Although scaling analyses have been performed for some of the severe. accident research experiments, there as yet is no systematic application of a scaling methodology that would allow confident extrapolation of observations made in intentionally scaled experiments to severe reactor accidents. The work to be done is to first. determine whether a scaling rationale and similarity criteria for severe accidents research can be developed to provide assurance that the correlations or models developed-can be-properly scaled to reactor condition. Second, if such a methodology is developed, a E systematic application of the same will be made to appropriate experiments and computer codes carried out under the SARP.
Task 1 Develon and Anniv a General Severe Accident Scalina Methodoloav Research Approach:
A review of scaling methods will be made. This review will be of scaling methods in general and methods used in the experiments of the SARP to date. From examination of the nature of the phenomena being studied, the experiments being conducted to study those phenomena, and the. review above, scaling methodology will be developed for the SARP. The efficacy of the methodology will be tested initially by applying it to the direct containment heating experiments that have been conducted in the SARP to date.
This task will not address directly whether the correct phenorena are being investigated in a particular experiment - that is done for each task throughout the SARP - but rather whether the experiment will indeed yield information about the phenomena being investigated relevant to behavior at the scale of an actual reactor.
i III-9 1
'1
. : :_i - > .-
- .e lUseables and Use, j To the1 degree that this work is successful, the SARP.will' have a scaling methodology that can be applied to the-experiments of the SARP to provide a more rigorous and systematic 111nk between experiment and! accident phenomena ,
than is present today. The application of the p, methodology will help in the. planning.and. conductof:the i experiments directed at'the-longer-term goals. :In _ the j
. . near-term,. application of'the methodology should help- -
- establish the level of confidence that.can-be placed in the present understanding of containment performance ~from scaled experiments and code analyses performed)to date.
Issue 2.- Decressurization and DCH p
'Present1 understanding ce the in-vessel progression of a PWR core degradationLaccident is. limited'but suggests the formation of melting-solidifying fronts as the molten core materials " candle idown"~into the: lower,Lcolder portions of the fuel bundles and solidify.-- This " frozen crust" of previously molten core material L C' -may' form'.in'a layer that inhibits steam flow from the lower plenum.
P -Unsupported fuel pellet stacks may collapse onto the crust and melt. As this process' proceeds, a crucible-shaped crust supporting a molten mass of core material may form in the central region of the core. Failure of this crust at a radial location-(" sidewall")
is expected from considering thermal convection of the melt within
/ -the crucible. In fact, this is what was concluded to have occurred at TMI-2 upon examination of the reactor core. Upon crust failure,
. molten core material would flow downward, possibly ablating through
~
the core shroud and flowing onto the flow distribution plate, ,
eventually reaching the lower head. However, very little III-10
1
. I experimental information on this procsos exists, and computer code predictions are at best uncertain. Hence, prior to dismissing crust failure on the bottom portion of the crucible (not a plausible failure mode), this issue will be assessed. The main difference between these two failure modes seems to be the amount e of molten material that pours into the lower head and is ultimately available to be released to the containment. Our current understanding indicates anywhere from around 25 to 60; percent of the core could relocate to the bottom head upon crucible failure.
-Because of the difficulty of conducting appropriate tests to better characterize these phenomena, questions associated with core relocation are likely to remain highly uncertain in the near-term, although modeling of these processes will continue under the SARP.
Thus, the question of lower head failure mode can be approached by considering a sufficiently broad range of debris quantities, composition, temperatures, and relocation rates. The TMI lower head inspection plays an important role in understanding the relocation process and the threat to the lower head integrity. On the other hand, depressurization of the RCS, either by structural
-failure of the pressure boundary (e.g., piping, steam generator tubes) resulting from natural circulation, or deliberate depressurization by operator action, may alleviate concerns over
~
DCH. In this regard, any analysis with respect to depressurization would be assessed in terms of the current state of operator training including information available to the operator (instrumentation) and it would include the dominant accident j sequences anticipated for a given plant.
III-11
4
-As part of the development of the Reactor Risk Reference Document (NUREG-1150), the DCH issue was presented to an expert panel that .
included experienced severe accident analysts.- Monte Carlo analysis was performed that integrated all of the pertinent phenomena. The resultant likelihood of containment failure was found to be small and the chance that containment would survive the L loads associated with DCH is much y eater than had been originally estimated and reported in the first draft of NUREG-1150. The experts based their judgments, including the likelihood of RCS failure by natural circulation leading to RCS depressurization, on the current body of research evidence.
' Consistent with the above, in the near-term the SARP will address the DCH issue by a two-pronged approach:
- 1. Before we can rigorously conclude on the b'enefit to intentionally depressurize the primary systems of PWRs, an improved understanding of the challenge to containment integrity must be obtained. The existing research efforts will be critically examined with respect to the following:
- a. What is the nature and character of the information most l likely to be provided on high-pressure core melt accidents in the next several years? Are continued commitment of time and resources likely to result in a commensurate advancement of understanding of DCH phenomena?
III-12
.. =.
b '.
.c
- 2. The efficacy of PWR depressurization will be examined. In particular,.both beneficial and detrimental aspects will be explored.- Review of relevant emergency procedures, dominant accident scenarios, and timing of required operator actions to
- depressurize also would be included in this examination.
i Using the results from 1 and 2 above, an updated estimate of the risk associated with high-pressure core melt accidents will be made,.along with an estimate of the risk. reduction that would be I
obtained.by intentional depressurization, taking into account any increase in risk associated with depressurization. Based on a careful-weighing of the net benefits of depressurization, a recommended course of action will be proposed.
Some of the key questions that the SARP will attempt to answer in the near-term in order to make the above determinations are:
- 1. What is the likelihood that the RCS will fail by natural circulation prior to lower head failure? If so, are these failures of much lesser concern to overall risk? (Task 2.2)
III-13 b
?'
,o
- 2. :Isithere a. low-pressure cutoff below which there isino DCH threat?- (Task 2.3)
- 3. If so,-will.this pressure be reached by natural circulation-induced failure of the RCS, or through. operator-action to depressurize, or'both? (Tasks 2.1 and 2.5)
>4.-. EIf operator action is.necessary,-is there time available for
-this action? Given that operator action is.to be taken what H are the' hardware and procedure specifications that would.
enable; successful-depressurization actions. (Tasks 2.3 and 2.4)
- 5. 'Are there adverse consequences to early depressurization? If so,'what are these and their potential significance in terms
-of, causing either earlier core damage or earlier containment failure? (Tasks 2.4 and 2.9) -
.6. What' is the nature of the DCH threat,.and what mechanisms (e.g., sprays) and configurations exist ex-vessel that will mitigate or eliminate it? (Tasks 2.4-2.8) l l
l 9
III-14 l
Logically an answer to this-last question should precede the rest.
However, even with DCH apparently presenting much less of a risk than previously believed,'there is not yet sufficient confidence in
)
l
, this belief.that it warrants delay in seeking answers from the other relevant tasks, especially in light of the 3-year horizon of the Integration Plan.
Task 2.1 Evaluate current research Drocram addressina hiah Dressure melt eiection challences to containments.
Research Approach:
Current research programs, both experimental and model development programs related to DCH and hydrogen production during high-pressure melt ejection (HPME), ~
will be reviewed and assessed to quantify the current level'of uncertainties to determine if any of the questions raised in the following tasks have already been answered. In addition, the assessment will determine the nature and character of information likely to result from these programs over the next 3 years. Progress made to date, current research topics, major uncertainties, and methods being used, both experimental and analytical, will be examined. This task will be carried out prior to the planning and conduct of any new' experiments or analyses.
Useables and Use:
Based upon the results of the above assessment, a
-judgment will be made as to whether a significant increase in understanding the threats to containment from high RCS pressure accidents will be achieved in the next 3 years. In particular, the question of whether a significant reduction in uncertainties will be achieved l
l only with a significant additional expenditure of time I and resources will be addressed.
III-15
7
- c. -
[
M Task 2.2- Assess the likelihood of RCS structural failure by L natural circulation.
Research ADoroach:
Calculations of RCS failure by natural circulation will !
be scrutinized and alternative models and assumptions
. examined. Comparisons of calculations with available experiments-(Westinghouse)'will be made to' identify potential weaknesses in both experiments and models, particularly with regard.to scaling. An assessment will l be made on why there were only minor natural circulation '
effects in the upper vessel structure.and none in the hot leg piping at TMI.
High RCS pressure PWR sequences will be selected by initiating event, characteristics of the early stages of L theLsequence, and potential for evolving to intermediate or low RCS pressure sequences (e.g., potential failure of a safety valve in a station blackout during core heat-up). This selection should include loop seal clearing and pump seal failures. Codes predicting energy redistribution due to natural circulation will be applied to these sequences and will be verified, validated, and examined for underlying assumptions that could affect the results. Analyses and comparisons of the selected sequences as to likelihood, location, and time of failure !
.of-the RCS boundary will be made. The effects of fission product. deposition and revaporization as an additional source of thermal energy during natural circulation will be considered.
i L Useables and'Use:
The result of this task will be an estimate of the likelihood of RCS failure from natural circulation in l high RCS pressure scenarios for typical primary system geometries. It is possible that this item will alleviate i
concern with the threat of direct containment heating.
l 1'
Task 2.3: Investigate the influence of cavity and conta'inment compartment structures on DCH, low-oressure cutoff for DCH, and hydrocen production from HPME.
Research Aeproach:
The scoping calculations performed for NUREG-1150 to l estimate the DCH load from the dispersal of melt into the containment correlated DCH load with melt quantity, l
ejection pressure, and ejection rate. The results'of this analysis identified the key unknowns in the prediction of DCH containment loads from melt dispersal l
III-16 l-
. _ . . _ __-i
4: ~4 b*
k that needito be determined by experiment, particularly, the effect of lower containment compartment geometries; the effect of water'(co-dispersal with the corium or added by= spray) on reducing the magnitude of DCH I -pressure;.and the production of hydrogen'by HPME.
Experiments at different scales (using existing facilities insofar as possible) with high temperature, thermitically generated melt-and steam as a pressurizing fluid will be performed. Measured quantities will be identified based on'the scaling analysis of Task 1.
Variables believed at this time to affect the magnitude of the challenge to:the containment are.nelt mass expelled from the cavity and its distribution within the containment, containment' atmosphere pressure' and temperature, melt metal content of the dispersed melt heat transfer. rates to structures, .and hydrogen production in the dispersing melt-steam mixture, and its transport and combustion in the containment.
Useables and Use:
Direct comparisons of containment pressurization observed in tests at different scales (existing facilities) provides insight on the dominant factors affecting scale up of the results. The data may be used to estimate melt entrainment as a function of quantity of melt and
>- ejection rate. The melt entrainment will then be used to calculate at what pressure DCH from core ejection does not threaten the containment integrity (low-pressure cutoff).
Research Annroach:
The overall framework for analyses should be based on a comprehensive treatment of representative scenarios, including possible perturbations by operator actions i
l (based on the current state of operator training, L
Emergency Procedures or Functional Restoration Guidelines, and timing for when the operator is required to depressurize) and equipment failures that may have a bearing on the issues addressed here. The research
'should consider, but not be limited to, the normally available power-operated relief valves (PORVs) as a depressurization device. One such strategy is to decrease the system pressure sufficiently to allow the accumulators to dump prior to significant cladding i oxidation, so that gross fuel degradation is approached at a pressure that will not threaten the containment III-17
V a .
p :5, s. ; w
' b; I ~d
-g a
@ integrity should the vessel' lower head. fail. The approach is essentially analytic, using thermal '
-hydraulic. codes such as RELAP5 or TRAC for the'most part, ,
q although selective use of core degradation codes may also- .j prove useful. The use of models in these codes that'have- 3 not been.testedlin circumstances peculiar to this l I
application,'e.g., reflooding of'a highly overheated core would be scrutinized to facilitate judgment as to the' meaningfulness of the results. :The research will examine-
~
both" beneficial and detrimental effects of early intentional-depressurization.
Useables and Use:
.The result of this-work' vill be a defined relationship between depressurization. strategies and: final' system pressure 1at the time of gross fuel degradation, or the time of lower head failure, and an identification of any detrimental; effects of intentional early RCS' L
.depressurization.
. Task 2.5: Determine the modelof bottom head failure of the reactor Dressure' vessel in a hiah RCS Dressure secuence.
-Research Acoroach '
-Assess and determine the likely modes of PWR pressure vessel _ bottom head failure'considering the reasonable range (Task 2.6) of quantities, composition, and timings of-core melt arriving on the bottom head. From the above assessment,-identify those experiments that may need to be1 conducted to determine the values of parameters that are-keys to1the assessments and/or confirm those assessments. Application of the scaling methodology (Task 1) will be made.in devising and interpreting the results of'the' experiments.
The TMI lower head inspection plays an important role in understanding why vessel failure did not occur for this accident. It is not sufficient to look for damage to the lower head. It is essential that the debris on the lower head be carefully characterized in order to interpret the observed condition of the lower head.
- Useables and Use:
An understanding of the sensitivity of reactor pressure vessel (RPV) lower head failure mode to the quantity, ~
composition, and tining of arrival of molten corium on the lower head. This understanding may be quantified in l:
III-18 g
l'
1 q
^
the form of system level codes that predict the mode and- ,
timing of reactor vessel failure in a manner consistent with the predicted melt progression and thermal-mechanical loading of the RPV. The results.of this task will contribute to the work of Task 2.3.
Task 2.6: Determine the likely rance of auantity, composition, and timina of molten core material arrivina on the bottom head of a PWR durina a core melt accident at hiah and l .
intermediate RCS oressures, j 1
Research Acoroach:
. Review.and assess existing experiments and associated analyses to ensure that the analysis tools represent the important phenomena. A series of core degradation and f ,
melt relocation calculations will then be made using L boundary. conditions and initial conditions that span the l range of. core melt sequences. For each set of conditions I considered, the calculations should yield the quantity, composition (proportion of metals and oxides),
l temperatures, and timing (rate of arrival) of melt l reaching the bottom head. Based on these calculations,
=bestiestimates will.be made on the characteristics of i
melt reaching the bottom head of a PWR in high and medium RCS pressure core melt accidents for assessing DCH. Low RCS pressure' conditions will be used in evaluating the potential effects of depressurization, including the effect of molten fuel-coolant interaction on core melt progression.
Useables and Use:
The result of this work will be the most likely characteristics of melt to be considered in addressing PWR bottom head failure (Task 2.5).
Task 2.7: Use the results obtained in Tasks 2.3 to unarade DCH models.
Research Acoroach:
. This task should proceed only if the results from the scaling program (Task 1) and Task 2.3 indicate that further analytical developments are necessary for calculation of DCH loading and closure of this issue.
Analysis may be used for extracting detailed information from experiments that can then be used to develop models e
that suffici'ntly represent plant-specific geometries.
The suitability of the models will be affirmed by r
III-19 1
9 'e
^
the form of-system level codes that predict the mode and I timing of reactor vessel failure in a manner consistent
.with the' predicted-melt progression and thennal-mechanical loading of the RPV. The results of this task wi'11 contribute to the work of Task 2.3.
Task 2.6: Determine the likelv rance of auantity, composition, and timina of molten core material,arrivina on the bottom ,
head of a PWR durina a core melt accident at hiah and '!
intermediate RCS Dressures.
Research ADoroach:
Review and assess existing experiments.and associated analyses to ensure that the analysis tools represent the important phenomena. A' series of core degradation and melt relocation calculations will then.be made using boundary. conditions and initial conditions that span the range of core melt sequences. For each set of conditions considered, the calculations should yield the' quantity, composition (proportion of metals and oxides),
temperatures, and timing (rate of arrival) of melt reaching the bottom head. Based on these calculations, best estimates will be made on the characteristics of melt reaching the bottom head of a PWR in high and medium RCS pressure core melt accidents for assessing DCH. Low RCS pressure conditions will be used in evaluating the potential effects of depressurization, including the effect of molten fuel-coolant interaction on core melt progression.
Useables and Use:
The result of this work will be the most likely
-characteristics of melt to be considered in addressing PWR bottom head failure (Task 2.5).
o Task 2.7: Use the results obtained in Tasks 2.3 to uncrade DCH models.
L Research Acoroach:
- This task should proceed only if the results from the scaling program (Task 1) and Task 2.3 indicate that further analytical developments are necessary for calculation of DCH loading and closure of this is,ue.
Analysis may be used for extracting detailed information from experiments that can then be used to develop models l that sufficiently represent plant-specific geometries. l The suitability of the models will be affirmed by
\
III-19 l l
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.c .-
~
comparisons of calculations to the results of experiments. The developed DCH models can then be used to predict containment pressure due to direct heating -
phenomena in reactor accidents as well as the mitigation of these phenomena by containment pressure control devices and~ procedures.
Useables and Use:
This work will result in DCH models that will be applicable to the evaluations of containment performance, the review of individual plafat examination submittals, and the assessment of the efficacy of depressurization and containment pressure control strategies on the i magnitude of DCH containment loading.
Task 2.8 Depressurization cost / benefit analysis.
Research Aceroach:
The results'of Tasks 2.1 through 2.7, and the results of l Tasks 4.1 through 4.3 dealing with fuel-coolant interactions at low system pressures, will be considered in performing analyses of intentional PWR system ,
depressurization as a mechanism to avoid DCH. The analyses.will be focused on assessing the efficacy of depressurization to bring about a net benefit to safety for important core melt scenarios (e.g., station blackout and high pressure conditions associated with loss of feedwater, and small LOCA scenarios. The benefit is the estimated reduction in risk of early containment failure from high-pressure ejection of molten core materials into the containment. Other' factors to be considered include equipment costs needed for intentional PWR system depressurization, and also any detrimental effects that may result from intentional depressurization. It is recognized that the results of Task 2.2 (Natural Circulation) may indicate that the RCS will fail and depressurize without operator action. If such is the case, whether to continue this task will be reviewed.
Useables and Use: ,
J The result of this task will provide additional bases upon which to recommend for or against requiring depressurization of PWRs during core melt accidents to avoid DCH.
l III-20 l
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I:
Issue 3 - BWR Mark I Containment Shell Meltthrouch An accident sequence leading to early containment failure has been postulated for BWR Mark I containments. This sequence involves the deposition of molten core onto the drywell concrete floor at the time of vessel breach, the subsequent spread of the melt to, and its contact with, the drywell steel wall that is the containment boundary, ultimately causing failure of the wall. Some experimental work has saggested that the addition of water in the drywell could provide a mitigative effect by causing fragmentation and quenching of the melt before it reaches the wall or by providing significant cooling of the melt layer. The heat transfer processes from the melt to the wall, the melt-concrete inter-
. actions, and the heat removal processes from the melt and from the wall are not completely understood, although some work has been done at Sandia National Laboratories on heat transfer from steel melts to steel structures and on melt-concrete interactions. There have also been experiments on bubbling heat transfer and on melt spreading at Brookhaven National Laboratory, using simulants, and experiments related to melt spreading and quenching have been conducted by Fauske and Associated Inc. In addressing this issue
- in the near-tern, the literature will be carefully reviewed to clearly identify the critical questions to be answered with respect to Mark I shell failure and the capability of the existing data to answer those questions before any new research efforts are begun.
III-21
S
. i one such question that has been identified is the sensitivity of L
the time and mode of reactor vessel lower head breach to the quantity, composition, and timing of arrival of molten core on the j i
bottom head of a BWR. The core melt of a BWR may proceed much differently from that of a PWR. For example, one possibility that has'been suggested is that the channel box and control blade geometry of a BWR suggests that stable crusts supporting the melt may not form (as in TMI) and that molten fuel will more or less
- continuously flow onto the lower head. Further, the lower head volume of a BWR is larger than that of a PWR and is densely occupied by the control rod drive structures, and this may result in a different failure process (e.g., less coherent failure) of the bottom head from that of a PWR.
It has been postulated that a BWR core melt progression proceeds by
. gradually accumulating fully quenched debris in the spaces between the control rod guides, in the lower head. Another postulated progression is similar to the TMI-2 accident with melt relocating into the water-filled lower head. In either case, following dryout, debris heatup will be accompanied by lower head heatup, and eventual failure of the bottom head. At this time, the composition and characteristics of the debris arriving on the drywell floor beneath the reactor vessel is uncertain. The debris could range from mostly solid oxidic with low superheat to mostly molten metallic with high superheat. The nature of the debris exiting the ,
failed bottom head of the BWR can determine the subsequent course i
III-22
]
of the accident. However, the bottom head itself may play a L
'significant role in determining the characteristics of that debris.
In light of.the above, the SARP will address the questions of how sensitive (or insensitive) to the quantity, composition, temperature, and timing of melt arrival is the failure mode of the bottom head of a BWR and to what extent are the hydrodynamic and thermal properties of melt arriving on the drywell floor determined by the mode of bottom head failure.
The BWR Mark I containment shell meltthrough issue, like the DCH issue,-was presented to a panel of experienced severe accident analysts who based their judgment on the current body of evidence.
The panel was equally divided as to whether failure would occur or would not. When the drywell floor is water covered, the exp,erts were more confident that failure would be avoided. Similarly, when the floor is dry and superheat is high, the experts.had less confidence tnat containment shell would remain intact. The panel concluded that the degree of belief regarding containment shell meltthrough is 0.33 or 0.87; the lowest failure probability corresponds to cases in which water is assumed to cover the drywell floor; and the highest corresponds to cases in which the drywell
/ floor is dry and debris flow rate, debris superheat, and debris unoxidized metal content are all high. Nevertheless, it was generally agreed that the composition and temperature of the debris i
exiting the reactor pressure vessel and the presence of water on the drywell floor are important variables in estimates of III-23
containment-shell meltthrough probability. This issus is of course, of interest where venting might-be accomplished after a .
severe _ core damage accident but it does not bear on'the recognized efficacy of venting prior to such an accident in order to prevent the same.
-Specific. questions that the SARP will attempt to answer in the near-term are the following:
1.: What is the relationship of the BWR bottom head. failure mode to variations in quantity, composition, temperature, and timing of arrival of the melt on the bottom head? (Task 3.1)
- 2. What is theLeffect of water-on the drywell floor when the melt
. pours out from the pressure vessel? (Task 3.2)
- 3. How does the answer to the above question depend on the
~ initial conditions (melt ejection rate, melt superheat, melt composition, initial presence'or absence of water) and water addition rate? (Task 3.2)
- 4. Under what conditions would the crust that forms at initial contact between the melt and the shell be stable and for how long? What is the expected rate of heat transfer between the core melt materials and the shell for various melt conditions?
(Task 3.3)
III-24 l
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sc1 , .. 1
. I 4 1 In addition to addressing the questions related to attack of the i
Mark I containment shell, the results of the following tasks will j 1
provide initial conditions for a variety of ex-vessel' phenomena that depend on the quantities and dynamics of spreading of the released corium.
Task 3.1: Investigate debris relocation ohenomena into the lower plenum of a BWR, includina the failure mode of the core clate.
Research Acoroach: ;
Review and assess existing BRR core relocation experiments and associated analyses. Analyze core plate and RPV failure mode and timing assuming a range of quantities, composition, and timing of melt. Based on these analyses, identify the key unknowns that can only be determined by further experiments,-and assess the feasibility of those experiments. If such experiments are determined to be both necessary and feasible, design and conduct properly scaled out-of-pile tests in which simulated molten core debris is poured onto a scaled core plate or bottom head to verify the failure mode as a function of the nature of the debris (quantity, composition, and timing). The results of this task, including that for any additional experiments carried out under this task will be used to develop a model to predict lower head failure. All the analyses and experiments should assume expected degraded BWR core conditions (e.g., water in the lower plenum at the time ;
of melt arrival on the core plate and ADS operation.)
Useables and Use:
This work will address the expected mode, timing, and temperature of core plate and bottom head failure and passage of molten core debris through core plate and bottom head for use in severe accident analyses. The i
results of this task will provide crucial initial conditions for a variety of ex-vessel phenomena that depend on the quantities and spreading characteristics of released corium.
III-25
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4 ii . .
[ . Task 3.2. Determine the effect of water on melt spreadina, melt
.coolina, and melt-concrete ~ interactions.
f Research Acoroach:
The results of melt spreading, cooling, and core-concrete interaction experiments will be reviewed to determine the l extent to which a consistent picture of critical variables and conditions can be drawn. Among the questions to be considered are the role of superheat, the effect of unexidized Zr in the melt, crust formation and stability, the generation of aerosols and noncondensible gases _from core-concrete interactions, and the effect on L the behavior of the melt of the presence of water--both water.beneath the RPV upon bottom head failure and water added~ atop the melt following the arrival onto_a previously dry floor. Consistent with the results from the scaling program (Task 1), calculations will be conducted either to confirm the conclusions of the review (or clarify issues that were concluded to be ambiguous) and define those near-term experiments that can be conducted to resolve ambiguities of data.
Useables and Use:
This work will produce refined estimates of melt spreading, hydrogen and other noncondensible gas generation from melt-concrete interactions, and the effect of water beneath the RPV. Of equal importance, however,-this task will-bring to focus those questions that-may only be addressable by experiment over the long-term since this task goes to some of the basic chemical and process effects of high temperature interactions of the core-concrete materials.
Task 3.3 Determine the conditions under which the interactions between the screadina melt and the Mark I drvwell shell will lead to containment shell failure.
Research Acoroach:
1 Experimental results of heat transfer from steel melts to steel structures will be examined, to determine their applicability to reactor accident conditions. In particular, questions such as the effect of the inclination of the shell to the drywell floor and the effect of concrete decomposition gas agitation of the melt will be addressed. Experiments will be devised and conducted to determine in the near-term whether a Unless stable crust can form and persist at the drywell shell.
i III-26
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_ - _ . - - .__ ---_ -- __ -_E
p V
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otherwise. indicated from the results of Task 3.2, this-task will' assume the presence of water in the drywell, The staff recognizes that it may well be that this task might not be completed in the near-term, however, available insights:and data will be used in the closure process.
The presence of Zi metal in the melt arriving at the shell depends on the extent of.its oxidation both b in-vessel and while the melt is spreading over the drywell floor.- If-it is determined that significant amounts of Zr remain unoxidized before contact with the shell, then the impact of the presence of Zr will be determined.
Useables and Use:
The results from this task will enable the staff to determine the conditions under which the shell of a Mark I containment may fail, and, in particular, determine the proximate melt amount, composition, and superheat necessary for failure.
. Issue 4 - Addina Water to a Dearaded Core There is little doubt that in a severe accident situation the priteery.effortsiof-the operators will be directed toward making water available to the reactor vessel. An important question.that needs to be considered in view of such likely efforts is what are the likely consequences of those efforts. Given the uncertainties i in core melt phenomena, the uncertainty associated with the operator's knowledge of the condition and location of the core during an actual severe accident, and given the intuitive drive to j put water onto the core in the event of an accident, it is not l likely that an operator ever would be told not to put water into l the reactor vessel should water become available during the course of an accident. However, along with the potential benefit of i III-27 1 \
_ _ _ ______________________a
achieving a stable, coolable configuration, restoring water to a core that has been severely damaged, can have effects of which the j operator should be aware. The operator also should maintain
)
cognizance of possible symptoms and response of the plant to adding i l
water in such circumstances (e.g., molten core-coolant interaction,
, increased hydrogen generation, it. creased containment pressure).
A related question that also needs exploration is what are the circumstances in which a grossly degraded core can be prevented from melting its way through the vessel lower head. Although important parameters can be identified as water availability, system pressure, and debris configuration, it is not likely that complete resolution to this issue will be achieved in the near -
term. Nonetheless, there are questions related to the addition of water to a degraded core that can be addressed in the near-term that will provide some insights into the longer-term issues of accident management as well as improve understanding of low RCS pressure core melt accident phenomena. The questions to be addressed are the following. j
- 1. What are the amounts and rates of hydrogen and steam
,I generation during the reflooding of a degraded core end during relocation of the melt onto the bottom head? (Task 4.1) 9 III-28
-8 4 IJ
- 2. What is the potential that reflooding a severely damaged BWR core will result in a recriticality? What will be the effects as opposed to not adding water? (Task 4.3)
Determine the effect of water iniection on the'aeneration Task 4.1 of steam and hydrocen durina refloodina of a dearaded Core.
Research Acoroach:
Identify basic variables governing heat transfer and hydrodynamics of melt-water. interaction, including the effect of water injection on debris reconfiguration. The work would be oriented toward scoping the range of hydrogen. generation that can be produced during quenching, and assess the impact on containment-performance. In consideration of the results of the scaling. analysis (Task 1), establish whatever additional experiments may be needed or conduct analysis needed to determine the effect of adding water to a degraded core.
We have already performed a lot of TMI-2 analyses and will assess tl.ese results.
Additionally, a benchmarking exercise, using available models (to be selected), against the TMI-2 accident
- < beginning at 174 minutes (when core was reflooded by the start of a reactor coolant' pump and high-pressure injection) will be undertaken as part of this task.
Useables and Use:
This task will produce estimates on the amounts and rates of steam and hydrogen generation as a function of water addition to degraded core geometries. The effect of adding water to degraded cores has important application to accident management as well as improving the quantification of the resultant containment loads.
l Task 4.2: Stability of Melt-Suoportina Crusts.
Research Acoroach:
The formation of a crust (crucible) supporting a melt, the predicted failure of the crust, and the eventual relocation of the melt into the bottom of the reactor vessel depend strongly on the heat transfer coefficients at boundaries of the core melt crucible. The anticipated approach is to review and assess existing analysis for III-29 i
1
L e ,.
- I both the' growth-and demise of a melt-supporting crust and subsequently perform a set of calculations of heat transfer along. boundaries of the crust crucible, .
analyzing thermal and mechanical. crust stability as a
' function of' melt accretion. J Useables and Use. ' 1 The work will affirm whether the heat transfer coefficient and the code treatments of core behavior, including crucible formation and collapse and debris relocation into the lower plenum, are suitable.
Task 4.3: Investigate the possibility'and consequences of recriticality in dearaded BWR cores.
Research ADDroach:
There are two configurations'to be considered with regard to the recriticality problem. One is the possibility that standing fuel pellet stacks will become critical upon refiboding (with control rod materials previously melted out). The other is the possibility that core material will become critical upon relocation to the bottom head.
The approach is to first review and assess existing analyses of both the likelihood and consequences of recriticality of a damaged BWR core. Where deficient, additional calculations will be performed assuming the existence of critical masses for expected degraded BWR core geometries reflooded with unborated water.
Unpables and Use:
The products of this task would be an estimate of the likelihood and effects of recriticality on accident.
progression and the alternative design mitigative measures (e.g., minimum boron concentrations in the reflood water) that might eliminate the problem if one is
- found-to exist.
l III-30 l
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- Issue 5: Use and Status of Severe Accident Models (Codes)
As noted earlier, the confidence in the analyses and judgments that will be made inLimplementing the staff's Integration Plan will depend on how well the tools used in performing those analyses and making those judgments conform to the realities of a severe accident. The tools that have been developed under the SARP, the physical and phenomenological models embodied in compdter programs (codes), have had and continue to have a key role in furthering the objectives of severe accident research, viz., support of the U.S.
Nuclear Regulatory Commission's policy towards severe accidents.
First, and perhaps most critical, is the use of the codes in identifying the important sources of risk. In this capacity the impact of the codes is important since the results of code calculations support the evaluation of risk. The evaluation of r.isk, in turn, significantly determines both what is researched and the relative priority of what is researched. Second, through an iterative process, exercising the codes helps def.ine the experiments needed to improve the representation of physical phenomena by the codes. Third, the complex mechanistic codes provide a benchmark for less detailed but faster-running codes that
,' are used in applications that require extensive calculations, such as parametric studies for evaluating accident management strategies. This "two-tier" code strategy has been pursued because the use of the detailed mechanistic codes to evaluate entire accident sequences is costly, impractical and unnecessary.
III-31
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' ' .' - .Any larga coda davalopatnt program is always fecad with tha difficulty of determining when the codes are " good enough." That is, when have they achieved an' accuracy sufficient for the -
i application-for which they were intended? At this time, an acceptable level of accuracy has not been defined. As a result, _J l
the current code development program is continually iterative. '
Codes are developed and assessed against' selected experiments, and needed model improvements identified. Overlaid on this process is
.the identification of new phenomena considered important to typical plants that "must be" incorporated into the code.
In addition to the above, a problem that is not unique to the severe accident program relates to the number of codes that must be developed to have a complete analysis capability. If one phase of the accident is difficult to model, or has inherently large uncertainties or variabilities, or perhaps is best understood as stochastic processes with a distribution of potential outcomes, it
-is not appropriate to model'other phases of the sequence in greater detail, since both the overall variability and uncertainty will be driven by the least precise and least certain processes, respectively. Currently, this appears to be the situation with regard to late-phase core melt progression. The degree to which
~
the early phase is modeled, and those models tested and validated by experiment, must be tempered by the degree to which the late !
phase through melt relocation and vessel failure can be understood and modeled, and that understanding tested by experiment. I 1
III-32
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.s 'Thir5foro,-es part-of1the code.dsvalopasnt program, a esthod'must
- be. developed that constantly assesses the state of code development
,4 from_agnumberLof standpoints.in addition ~to,how well a particular .
' sequence.or facet of an accident can be modeled.' The following Lquestions willibe considered:
-How well'do the mechanistic models reflect the phenomena 1.
believed to be important to severe' accidents? (Are the correct phenomena being modeled?)
'2. How well does the; interactive program of code advancement and
' experimentation achieve the objective implied in 1 above?
- 3. Is the. level of detail in the codes appropriate to their use?
(Are some codes more detailed than needed, others not detailed
.enough?)
- 4. Which stages of an accident need to be modeled by detailed e
mechanistic codes",. coupled to adjacent stages?- -(While clearly there-isLan interdependency among stages of a severs core melt accident, a detailed mechanistic code coupling all stages of accident phenomena from the onset of core uncovery through
- containment failure is not needed to either understand or make regulatory judgments about severe accidents. Further, a detailed mechanistic model of all aspects of a severe accident may not be possible, lot alone necessary.)
III-33 l
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- 5. . Is tha level of precision needed for regulatory use bning considered in the code development program?
- 6 '. Is'the level of precision from a given code more than.needed
'and.is.it suitable to the expected overall levels of uncertainty from an integrated analysis package for applications as discussed above (e.g., accident management)?
L.
Answering the above questions is a major undertaking and is not likely to be achieved in the near-term. Nonetheless, some measure of confidence in the tools that will be used in achieving the near-term goals must-be made. Therefore, the following task will'be done. Although not a complete code assessment, this task is believed to be a workable compromise in the near-term and should provide sufficient information to enable the staff to gauge the appropriate level of confidence to be placed in calculations using the codes.
Task 5 . Code Documentation and Review.
Research Approach:
For each code developed under the SARP, the developer will be asked to supply the following information:
- a. Stages in a severe accident to which the code is I - intended to apply and the accident sequences in which those stages are found.
- b. Current capability of code. (Of'the intended accident stages and phenomena, which does the code now model, which does it not).
t III-34 1
1
c.- Limitations of the code. (Portions ~of accident
-stages and/or special; assumptions under which the
. code operates).
- d. Statement of.why the code needs to be or does not-need to be coupled sequentially to adjacent codes to yield sufficiently precise calculations.
- e. . The degree to which the relevance of the physics, as
. embodied in the code, to severe accidents has been tested by experiment.- -(This point questions not L simply the conformance of code to experiment, but experiment to severe accident.)
- f. What features (physics and chemistry) not presently in the code /model need to be incorporated to enable the code to be a reasonable representation of accident phenomena.
lE g. What quality assurance program was the code subjected to.
Useables and Use:
The above information will be reviewed and scrutinized by the staff, and an assessment will be made of each code as to its'present scale up capability and utility to support achievement of the near-term goals of the revised SARP.
~
Based on that assessment, recommendations as to use,of the present form of the code, further development, or abandonment will be made.
Criteria for judging the suitability of each code are being developed. These criteria will take into consideration the . intended end use of the codes and the expected user of the code. Deciding when code i development is-completed would require a subjective judgment and a sense of proportion between the level' of precision needed for application and the levels of uncertainty and variability in models of sev3re accident phenomena.
This latter task is applicable to both the near and long-term research as is Task 1 on similitude. The conduct of Task 1 and 5 is believed logical and necessary to the critical reexaminations of SARP needs. However, neither is intended to constrain closure of the near-term issues '
described above.
III-35
-IV.. MEETING Lona-Term GOALS The research described in this.section of the SARP plan consicts of work primarily directed at reducing uncertainties in the estimation of.the risk presented by severe accidents. This orientation-is consistent with the direction provided in " Review of Research on Uncertainties in Estimates of Source Terms from Severe Accidents in Nuclear Power Plants," NUREG/CR-4883, known as the "Kouts Report."
That direction suggested a broad base approach, investigating manifold phenomena with the common goal of reducing uncertainties.
Both the approach and the goal remain valid, as does the more specific direction in NUREG/CR-4883 that identifies individual technical questions as contributors to uncertainty in risk estimates. However, there are sources of new information that bear on how to order research and distribute resources among the topics of the SARP. These sources are the revised " Reactor Risk Reference i Document," NUREG-1150, and the guidance provided in SECY-88-147 that identifies accident management as a strategy that the NRC will pursue in dealing with severe accidents. In addition, the successes of the SARP itself over the years are a source of
'information providing a clearer picture of which phenomena are
. critical to understanding severe accidents, insight into what avenues of investigation are likely to be most fruitful, and which l
topics are likely never to be fully resolved in a reasonable time frame within realistic resources.
l IV-1
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Taking the above'together with the needed balance discussed in Section II, specific topics have been identified that characterize the SARP efforts directed toward achieving the long-term goals cited in Section II - providing an understanding of the range of l phenomena exhibited by severe accidents, including impacts of generic. accident management schemes, and developing improved methods for assessing the " severe-accident source term." These topics are the-following:
- 1. Review of the SARP approach to modeling severe accident phenomena
- 2. In-vessel core melt progression and hydrogen generation
- 3. Hydrogen Transport and Combustion
- 4. Fuel-coolant interaction
- 5. Molten core-concrete interactions
,' 6. Fission product behavior and transport !
(.
- 7. Fundamental data needs IV-2 l
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', It 10 not at all clear that dstalled mechanistic modsla can bn, or need be, developed for items 2 through 6. Some phenomena may only be tractable when treated statistically as a population of related events with differing outcomes. In recognition of this possibility, the first item listed is a. systematic review of the approach taken in the SARP to modeling severe accident phenomena.
Issue L1 - Modelina Severe Accidents.
Because of the difficulty in performing prototypic experiments and the variety of scenaries possible, substantial reliance must be placed on the development and validation of complex computer codes for analyzing severe accidents. A number of " mechanistic" codes (e.g., SCDAP/RELAP, MELPROG/ TRAC) have been developed for various stages in severe accidents, both in-vessel and ex-vessel, for both PWRs and BWRs. However, it appears that there are practical limits to the feasibility of deterministically modeling all aspects of severe accident behavior. Moreover, it is not clear that all aspects of severe accidents need to be mechanistically reflected in deterministic codes. The resolution of some severe accident regulatory issues may be achieved with bounding analyses alone.
Further, some phenomena may bc better addressed employing stochastic rather than deterministic techniques. Priorities for additional development can be determined through use of probabilistic risk assessment techniques. When risk is unsensitive to the uncertainty in the phenomena or variabilities in parameter values, there should be no need to attempt to reduce the uncertainty or variability further.
IV-3
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.g .
Using the " documentation" provided by Task 5 of Section III of this l' ,
plan as.a point of departure, the approach of developing
=
. mechanistic,-deterministic codes for each stage of severe acc idents will be reviewed. For each code development program / phenomenon I
.being addressed, a review of progress to date, magnitude of uncertainties remainir.g, and degree of validation will be made.
Based upon this information, a determination will be made as to whether an alternative approach might be appropriate for dealing with some phenomena.- The efficacy of the.SARP's "two-tier" code strategy (fast-running lumped parameter system codes benchmarked against a combination of individual stand-alone and coupled mechanistic, deterministic codes) also will be examined. This review will begin'in the beginning of FY 1990 and recommendations will be made atithe end of FY 1990.
The.NRC staff assessment of the modeling program will be assisted b'y various individuals outside the NRC and will be drawn from universities, national laboratories, industry personnel who are expert in accident phenomenology-(have appreciation for the purpose of the analysis and the uncertainties in the all parts of the analysis), regulatory concerns, and programming and numerical techniques. This effort would identify the strength or weakness of different modeling approaches, quantify the uncertainties in code models, and identify additional research, if any, where uncertainties and risk importance are significant. ,
IV-4
_ _- _ - __ _ _____-___-_ _ _ - _ - _ - _ _ = _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ - - - _ _ = _ _ _ _ - _ _ - - _ _ - - - - --
t 4 .
Issue L2 - In-vessel core melt Drocression and hydroaen aeneration.
In-vessel core melt progression concerns the state of the reactor core from the start of core uncovery to reactor vessel failure.
Included phenomena t.re thermal attack by the core debris upon the reactor structure and the reactor vessel, in-vessel hydrogen generation, in-vessel natural convection and heat transfer, and in-vessel steam generation. The details of core melt also determine rates and amounts of in-vessel fission product release and aerosol generation and much of the fission product and. aerosol transport (and retention) in the reactor coolant system. (Some aspects of this issue relating to fuel-coolant interaction were discussed in Section III.)
s In considering melt progression, BWRs and PWRs need to be treated I separately, mainly because of their different fuel assembly, control element, and lower plenum structures. For example, the effect of water injection on (1) the integrity of upper in-vessel structures, (2) the mode of core relocation,, (3) hydrogen generation upon core relocation, and (4) the mode of bottom head
. failure and the details of the release of core and structural debris into the drywell following failure are questions common to both PWRs and BWRs. However, the answers are quite distinct 1
arising from the physical and geometric differences between BWRs ,
and PWRs. Bounding calculations and few small-scale experiments I
may be continued to identify which phenomena are critical for each l IV-5
i' 'n
- l reactor typo cnd whether further reduction in uncartaintice and calculational variabilities related to melt progression and hydrogen generation are needed and achievable.
Areas in which work will continue are the examination of TMI-2 data i
for whatever-insights might be provided, the NRU fuel melt l experiments, and participation in the existing cooperative program 1
with the FRG at the CORA facility. New experiments that evolve in the near-term research may continue as well.
Issue L3 - Hydrocen Transport and Combustion The major concerns regarding hydrogen in LWRs are that the static or dynamic pressure loads from hydrogen combu'stion and detonatioh may breach containment integrity, or that safety-related equipment may be damaged as a result of either pressure loads or high temperatures. To assess the possible threat to containment and safety-related equipment, it is necessary to understand how p-hydrogen is transported and mixed within containment and to determine the likelihood of various modes of combustion. The I hydrogen behavior issues have been extensively investigated since L
1979, but several important areas of uncertainty remain:
(1) high-temperature /high steam concentration combustion and (2) deflagration-to-detonation transition t
l IV-6
4 9-
. 1 Research programs on these issues of combustion.have been sponsored
~
in the U.S. .by the NRC and the nuclear industry as well as in the -
international community. However, the combustion processes are sufficiently complex such that many aspects are still not well understood. 'The resulting uncertainties in the threat to containment integrity are unlikely to be reduced significantly by the existing research programs. The intent of the this research is to reduce these uncertainties, however in some cases reduced uncertainties are not required to make a near-term regulatory decision. Each category of uncertainty is discussed below.
Hich-Temperature Cohbustion The Zeldorich-von Neumann-Doering (ZND) chemical kinetics !
theoretical model developed under NRC sponsorship predicts that increasing temperature has a strong effect on the combustion and the detonation of off-stoichiometric hydrogen-air mixtures and on all hydrogen-air-steam mixtures. It is believed that the steam inerting effect is reduced greatly at elevated temperatures. There is a limited supporting data base for this phenomenon. Experiments are necessary to resolve the uncertainties associated with high l
temperatures and steam concentrations typical of those likely to be encountered in severe accident scenarios. It is also necessary to extend or develop more mechanistic models to predict, by either extrapolation or interpolation, the temperature sensitivity of hydrogen-air-steam mixtures that have not been tested.
IV-7
s The small-break LOJA and TMLB scenarios are two examples of high-temperature hydrogen-air-steam mixtures that may exist below the auto-ignition temperature (550 C minimum value for stoichiometric, hydrogen-air mixtures). There are two aspects of the high-temperature /high steam concentration combustion problem. The first is the injection of high-temperature hydrogen and steam mixtures I
that auto-ignite upon contact with pre-existing and premixed hydrogen-air-steam mixtures. The second aspect is the injection of hydrogen and steam mixtures at elevated temperatures that do not auto-ignite upon contact with pre-existing hydrogen-air-steant mixtures. This allows the possibility of a premixed condition to form and a subsequent deflagration or detonation. In both cases, the competition between chemical reaction rates and physical mixing rates will ultimately determine the ensuing c"ombustion mode. Data are needed to determine the hydrogen-air-steam flammability limits and steam inerting criterion at elevated temperatures if reliable predictions are to be made as to the lixelihood and the potential threat resulting from various combustion mode (s) for a wide range of accident conditions.
Therefore, the research approach is to determine hydrogen-air-steam
! flammability limits, volumetric oxidation (chemical reaction) rates, physical mixing rates, and the competition of these rates during the injection of high-temperature hydrogen and steam mixtures into cooler pre-existing and premixed hydrogen-air-steam mixtures.
IV-8 l
s 4
i A draft proposal is currently under consideration by the NRC that addresses feasibility; construction design, cost, and schedule; and ;
an experimental test plan. It is our current estimate that if a j facility is needed, it can be constructed by early FY 1990 and that experimental results can be generated by late FY 1991 or early FY 1992. Final data reduction and formal documentation should be available in late FY 1992.
Containment Loads for Detonations Deflagration-to-Detonation Transition (DDT)
Direct initiation of a detonation would require a ' concentrated high-energy source for insensitive, steam-dil-uted mixtures. This is not considered a credible mechanism of initiation by almost all researchers. However, it is possible to initiate a flame with a low-energy L-'rce such as a spark or a glowplug, and the subsequent propagation through orifices and around obstacles such as pipes can result in flame acceleration that culminates in a transition to detonation.
The possibility of DDT in realistic and prototypic containment geometries and conditions needs to be resolved. The uncertainty in this area has increased because of recent experimental and theoretical results that indicate an increased likelihood of IV-9
detonations at high temperatures and large steam fractions as l
discussed earlier. The current data base on flame acceleration and .,
i DDT suggests that the mixture composition, obstacles, and venting i are all important factors. These uncertainties result from a lack of experimental data on flame acceleration and DDT for conditions that include the effects of steam dilution, elevated temperature, large-scale and prototypical obstacle types, and spacing.
At present, no reliable model exists to predict or extrapolate DDT results from small-scale experiments to containment scale. Reactor safety studies and fundamental combustion research is being carried out in the Federal Republic of Germany and Canada. These data along with data generated for space shuttle application will be ,
applied to reducing the uncertainty associated with DDT as well as in assessing the potential threat of DDT to containment integrity.
Specifically, this data base will then serve as the basis to improve or develop correlations and models to allow extrapolation of experimental results to reactor scale and accident conditions.
Issue L4 -
E_uel-Coolant Interactions.
Molten fuel contacting water can give rise to a range of phenomena.
Very energetic steam explosions could result in early containment failure (alpha-mode failure). Less energetic interactions do not threaten containment directly but could change the course of the accident and the magnitude of the source term. Among these are IV-10
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - .~
l sudden coherent failure'of the vessel lower head, ex-vessel debris
' dispersal and steam generation pressure pulses. (Core debris dispersal','and steam and hydrogen generation in some BWR Mark II and Mark III containments also warrant further evaluation as to i
whether fuel-coolant mechanism can pose a threat to containment
. integrity.)
The research to be conducted will be selective, confirmatory in
. nature, and' focused to address some well-defined questions dealing with key parts of steam explosion phenomenology, such as degree of premixing, triggering, and fragmentation within the detonation wave of a steam explosion for various premixture conditions. It is.
clear that the program will necessitate limited scale experiments to test the predictive capability of calculations of premixing and C .the fragmentation rate of melt drops in the explosion zone of a propagating explosion. Further, the utility of the calculational tools to assess the effect of (1) reflood of a damaged core or debris bed in-vessel, and (2) fuel-coolant. interaction in a suppression pool will be examined.
In addition to providing final confirmation of the staff's position on the issue of the alpha mode of failure, this work also will provide analytical tools, additional data, and insights for use in evaluating the dynamics of molten fuel-coolant interactions in in-vessel and in various containment configurations. .
IV-11 t
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5.: Quantify 1the effects of scale distortion.
g 6. Quantify the effects associated with extrapolating correlation and/or models beyond their data base.
- 7. Relate in a systematic and unifyina manner-
- test facility design and operation-test data accuracy'and applicability
- correlations (or model) accuracy and applicability to:
computer code scale-up capability and applicability to-calculate the postulated scenario for a full-scale nuclear power plant.
I
- 8. Provide a quantifiable and traceable procedure for specifying and prioritizing future experiments if needed.
It should be noted that requirements 4 through 7 provide the
.information needed to quantify the code uncertainty to calculate the postulated severe accident scenarios.
A scaling; methodology that meets these requirements is outlined in Jwhat~follows.
B2-8 f'
L - - __-___ - - -_ - - - _
'l.* .
FIGURE 2.1 SEVERE ACCIDENT SCALING -
METHODOLOGY (SASM) 4 SPECIFY SCENARIO ELEMENT 1: i ESTABLISH EVENT TREE IDENTIFICATION AND F
SEVERE ACCIDENT PATH HEN MENA I
IDENTIFY AND RANK PHENOMENA (PIRT)
PERFORM A TOP-DOWN ELEMENT 2: SCALING ANALYSIS
, IDENTIFICATION AND SYSTEMATIZATION OF APPLY SCALING CRITERIA GROUP THEORY APPLY ELEMENT 3: FUZZY-SET THEORY OUANTIFICATION ELEMENT 4: APPLY APPLY APPLICATION TO TO MODELS AND EXPERIMENTS COMPUTER CODES ,
y . .
[-.- ..
L 4. Elements 1of SASM and Their Rationale The proposed Severe-Accident Scaling Methodology (SASM) consists of four primary elements as shown in Figure 2.1.
The first element: Identification and Ranking of Phenomena contains Steps 1-3. In this element, scenario modeling (and therefore scaling) requirements are identified by (a) opecifying the scenario (Step 1), (b) establishing the event tree and selecting the accident path (Step 2), and (c) identifying phenomena / processes along this path and ranking their importance (Step 3). Activities carried out in this element are designed to meet the first requirement set forth in the preceding section.
These activities are similar to those discussed and developed in References'2 and 3, that is, they:
- 1. Provide a comprehensive, physically based framework for analyzing an accident scenario. This is essential to a L scaling methodology that is systematic, auditable and traceable.
- 2. Decompose the scenario in elementary components and I provide a casual relationship approach. This is essential for identifying modeling (and therefore scaling) requirements and understanding the role of each component-(and therefore model) in the selected accident path.
B2-10 i i
?
4' Identify and rank processes and phenomena along the 3.
accident path. This is important to a scaling methodology that is not only systematic but also practical. The need for this screening process arises from the fact that it is not feasible either to design an experiment or to develop a code that will scale properly all processes occurring during an accident. What is needed, however, is to ensure that physical and chemical processes important to the evolution of an accident are properly scaled.
- 4. Identify and prioritize research activities directed at resolving potential safety concerns related to an accident scenario. This is essential.to an expeditious closure of a safety issue. i The second element: Identification and Systematization of Scaling l Criteria contains steps 4 and 5. In this element, scaling criteria are (a) identified by performing a top-down scaling analysis (Step
- 4) and (b) systematized (Step 5) by applying group theory methods ;
.' discussed in References 4, 5, and 6. Activities carried in this element are designed to satisfy the second and third requirement listed in the preceding section, that is, they:
B2-11
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- 1. -structure-the scaling process to follow the physically r
. based,-hierarchical 1 approach detailed in Element 1. Such E ,
a structure is essential as it provides not only the i rationale for establishing scaling and similarity
~
I . criteria but it' ensures also that processes important-to l Lthe scenario are taken into-account in the similitude y ^
. analysis.
L 2. EProvide a structure and procedure that starts'from a L- global, top-down view point and introduces complexity.and t.
, detail at each lower level. This is important for l.. conducting comprehensive reviews of facility design, test specifications, and results. 1 l
j l
- 3. Provide a systematic and traceable approach to derive.and i select similarity parameters. This minimizes the arbitrariness, that is, the ad hoc approach used so often
'in' facility design and test specifications.
1 4.- Provide a method that which can yield similarity criteria for processes that can limit the operational range of a
..- system, l'
- V
.B2-12
_-____m__m m._-.____.___._-_m.______-____----__.___._______m __
I. ,
l0 The third element: ouantification contains Step 6, with activities designed to meet requirements 4 through 6, that is, to quantify the effects associated with scale distortion and/or with extrapolating correlations beyond their data base. Several methods can be used to achieve these objectives. One among them based on the fuzzy sets theory of Zadeh (Refs. 7 and 8) appears very promising in view of its successful application by Kubic and Stein (Refs. 9 and 10) to be a problem concerned with quantifying model uncertainties and system similarity.
The fourth element: Anolication which consists of Step 7, is designed to satisfy requirements 7 and 8, that is, to provide the same methodology that can be used to:
- 1. Design and operate test facilities.
- 2. Evaluate test daia accuracy and/or applicability.
- 3. Evaluate correlations and/or model accuracy and applicability.
.' 4. Evaluate computer code scale-up capability and applicabil'L:.
B2-13 1
... . o i
4, It should be noted that treating experiments and code model/ correlation' development in a systematic and unifying manner ,
(by applying the same scaling methodology to both activities) is a
. prerequisite for-quantifying efficiently and more~ accurately code uncertainties to calculate the postulated severe accident scenarios-in a full-scale nuclear power plant.
- 5. ADolication and Demonstration I l
A program has been initiated at the Brookhaven National Laboratory with the objectives to:
- 1. Develop a scaling methodology (SASM) that meets the eight
-requirements listed-above, and
- 2. Demonstrate the methodology by applying it to the DCH problem.
e Furthermore, a Technical Program Group (TPG) has been formed to assist the staff and provide guidance to this work. The members
-were selected on the basis of their:
F
- 1. Knowledge of severe accident phenomena and issues, and/or ,
I l
l'
- 2. Internationally recognized expertness in modeling and scaling of complex systems and phenomena.
B2-14
'l. .,
I; 3,
L In-order.to provide input from a broad spectrum of technical sources,;the composition of the TPG was specifically designed to L: . include technical talent from universities, national laboratories, and industry. .
This work that is, the development and' demonstration of SASM, is i: expected to be completed by the end of December 1989.
I
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9 t
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f' B2-15 d.. i
, FIGURE 2.1 j SEVERE ACCIDENT SCALING I METHODOLOGY (SASM?
I SPECIFY SCENARIO ELEMENT 1: i IDE TIFiCATION AND ESTABLISH EVENT TREE N ME SEVERE ACCIDENT PATH I
- DENTIFY AND RANK PHENOMENA (PIRT)
PERFORM A TOP-DOWN ELEMENT 2: SCALING ANALYSIS IDENTIFICATION AND SYSTEMATIZATION OF APPLY SCALING CRITERIA GROUP THEORY APPLY ELEMENT 3: FUZZY-SET THEORY QUANTIFICATION ELEMENT 4: APPLY APPLY APPLICATION TO TO MODELS AND EXPERIMENTS COMP'JTER CODES
,' .f ) . j,
-r,,,. -
g.
.. 6. References for Accendix B i- a
- 1. N..Zuber, "A Hierarchical-Approach.to Scaling and a Severe Accident Scaling Methodology," NUREG Report to be published.
h'
- 2. T. Theofanous, " Dealing with Phenomenological Uncertainties in Severe Accident Assessments and Prrbabilistic Risk Analyses,"
t.
LProceedinas of the Third Internationtil Toolcal Meetino on Nuclear Power Plant ~ Therma-Hydraulics and Operation, Seoul, I
-Korea, November 14-17, 1988. -
- 3. " Quantifying. Reactor Safety Margins: Application of Code Scaling, Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss of Coolant Accident,"
Technical Program Group, NUREG/CR-5249, 1989.
- 4. H. A..Becker, "Dimensionless Parameters, Theory and Methodology," Acol. Sc. Publ., London 1976.
S. S. T. Kline, " Similitude and Approximation Theory," Springer, N. Y., 1986 I
i-l
- 6. A. Laszlo, " Systematization of Dimensionless Quantities by 1 i
Group. Theory," Int, of the Heat and Mass Transfer, Vol. 7, ;
p pp. 423-430, 1964.
B2-16 i-L.
'i Zadeh', " Fuzzy Sets," Information and control, Vol. 8,
.7... ~ L..A.
o pp. 338-353, 1965.
!' 8. L. A. Zadeh, " Similarity Relations and Fuzzy Orderings," Inf.
Sci., Vol. 3, pp. 177-200, 1971.
[.
- 9. 'W. L'. Kubic, and F. P. Stein, " Concept. of System Similarity in Chemical Engineering," AICHE Y., Vol. 33, pp. 1986-1987, 1987.
3
- j. -10. W. .L. Kubic, and F. P. Stein, "The Application of Fuzzy Set
! Theory to Uncertainty in Physical Property Models," Fluid Phase Eauilibria,-Vol. 30, pp. 111-118, 1986.
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=5 ADVANCED COPY TO: The Public Document Room DATE: S/4,/94 j FROM: SECf Correspondence & Records Branch ll E
!l Attached are copies of a Commission meeting transcript and related meeting
- j. document (s). They are being forwarded for entry on the Daily Accession List and j' placement in the Public Document Room. No other distribution is requested or gl required.
Meeting
Title:
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Meeting Date: 7bt4 A /184 Open Closed i!
ll Item Description *: Copies ,
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