ML20209H428
| ML20209H428 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1986 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-1080, NUREG-1080-V03, NUREG-1080-V3, NUDOCS 8609150249 | |
| Download: ML20209H428 (115) | |
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NUREG-1080 Vol. 3 Long-Range Research Plan FY 1987-FY 1991 U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include N RC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal egencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations,and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
NUREG-1000 Vol. 3 Long-Range Research Plan I
FY 1987-FY 1991 Minuscript Completed: August 1996 D:ta Published: August 1986 Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission W:shington, DC 20666
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U.S. Nuclear Regulatory Commission Previous Reports in Series NUREG-0740, "Long-Range Research Plan, FY 1983-1987," Office of Nuclear Regulatory Research, published March 1981.
NUREG-0784, "Long-Range Research Plan, FY 1984-1988," Office of Nuclear Regulatory Research, published August 1982.
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NUREG-0961, "Long-Range Research Plan, FY 1984-1988," Office of Nuclear Regulatory Research, published April 1983.
I NUREG-1080, Vol. 1, "Long-Range Research Plan, FY 1985-1989," Office of Nuclear Regulatory Research, published September 1984.
NUREG-1080, Vol. 2, "Long-Range Research Plan, FY 1986-1990," Office of Nuclear Regulatory Research, published August 1985.
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TABLE OF CONTENTS Page PREFACE...............................................................
v INTRODUCTION..........................................................
ix 1.
OPERATING REACTOR INSPECTION, MAINTENANCE, AND REPAIR............
1-1 1.1 Reactor Vessels.............................................
1-1
- 1. 2 Steam Genarators............................................
1-6
- 1. 3 Piping......................................................
1-7
- 1. 4 Electrical and Mechanical Components........................
1-11 1.5 Nondestructive Examination..................................
1-16 2.
EQUIPMENT QUALIFICATION..........................................
2-1 2.1 Qualification of Mechanical Equipment (Environmental).......
2-1 2.2 Dynamic Qualification of Equipment..........................
2-2 3.
SEISMIC RESEARCH.................................................
3-1 3.1 Seismic Hazard..............................................
3-1
.3.2 Seismic Risk................................................
3-4 3.3 Seismic Margins.............................................
3-8 4.
REACT 1R UPERATIONS ANO RISK......................................
4-1 4.1 Risk Methodology and Data Developmenc.......................
4-1 4.2 Operational Safety Reliability..............................
4-4 4.3 Severe Accident Risk Assessment.............................
4-6 5.
THERMAL-HYDRAUL IC TRANSIENTS.....................................
5-1 5.1 Separate-Effect Experiments and M Mel Development...........
5-2 5.2 Integral Systems Experiments................................
5-5 5.3 Code Assessment and Application..................
5-10 5.4 Nuclear P lant Analyze r and Data Bank.......................
5-13 5.5 Technical Integration Center..................
5-16 6.
SEVEkE ACCIDENTS..............................................
6-1 6.1 Severe Accident Sequence Analysis.................
6-1 6.2 in-Vessel Core-Mel t Progression............................
6-4 6.3 Hydrogen Generation and Control..........................
6-6 6.4 fuel-Structure Interaction...............................
6-8 6.5 Containment Analysis..............
6-10 i
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TABLE OF CONTENTS (Continued)
Page 6.6 Fission Product Release, Behavior, and Transport............
6-12 6.7 Containment Failure Mode....................................
6-14 6.8 Fission Product Contro1.....................................
6-17 6.9 Severe Accident Risk and Risk-Reduction Analysis............
6-18 7.
RADIATION PROTECTION AND HEALTH EFFECTS..........................
7-1 7.1 Metabolism and Internal Dosimetry...........................
7-1 7.2 Health Effects Standards Development........................
7-1 7.3 Occupational Radiation Protection...........................
7-2 8.
WASTE MANAGEMENT.................................................
8-1 8.1 High-Level Waste............................................
8-1 8.2 Low-Level Waste.............................................
8-5 GL0SSARY..............................................................
G-1 iv
1 PREFACE This Long-Range Research Plan (LRRP) is intended to provide the Commission with a framework for planning research relevant to currer.t regulatory objectives or to future needs.
It was developed in accordance with the Commission policy and planning guidance presented below.
P_olicy*
1.
The research program is an essential element in many of the Agency's missions.
The research program should:
provide the technical basis for rulemaking and regulatory decisions; support licensing and inspection activities; assess the feasibility and effectiveness of safety improvements; and increase our understanding of phenomena for which analytical methods are needed.
2.
There should be continued emphasis on using research results in the regulatory process and on obtaining results that are useful therein.
The staff should be alert to research which shows that we ought to change our regulations.
NRC regulations should be changed when research shows them to be either too stringent or not stringent enough to adequately protect the public health and safety.
3.
The NRC will assess the ability of the containments to prevent acci-dental releases of radioactivity under postulated accident conditions so that the public can be assured that, in the event of an accident, radioactivity is contained within the facility. The Commission will continue to develop containment performance criteria.
4.
NRC will continue to maintain a long-range research plan which is consistent with the agency's mandate and directed toward areas of importance to the licensing and inspection processes. The long-range research plan and the NRC's Five-Year Plan will be consistent with each other.
Research resources should be allocated to support a balanced program between research to reinforce or revise the current regulatory base and conceptual research for improved reactor safety, waste management, and other licensed activities.
The major research comitment for NRC research efforts will be light water reactor safety.
The priority for research should be assigned in accordance with its safety significance.
5.
Joint or coordinated research programs with industry groups, other government agencies and foreign groups should be pursued when possible, both to expand the technical breadth provided to projects and to maximize the benefit to be derived from limited resources.
"This is the section on research in NUREG-0885, Issue 5, " Policy and Planning Guidance, 1986."
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l' a i-l Due consideration should be given to questions of conflict of interest when contemplating joint or coordinated research with industry.
Research results should be exwnined through the use of strong, effective peer review groups.
6.
The Commission has decided to proceed expeditiously with further characterization of radioactive source terms.
Planning Guidance 1.
The staff should continue their efforts to investigate the feasi-bility of containment performance criteria that will ensure effective containment performance.
If such criteria are shown to be feasibic, and practical, a set of such criteria should be developed.
o 2.
The radioactive source terms and their risk importance should be better characterized by a systematic analysis of accident sequences, containment performance and the release and traasport of radioactivity.
3.
The staff will coordinate a research effort to identify measures which can be taken to correct deficiencies attributable to aging an(
irradiation and to reduce safety risks inherent to degraded equipment.
4.
In its safety systems evaluation, including postulated accident sequences, the NRC will conduct research on complex operational and thermal-hydraul' transients.
l.
I 5.
The NRC will continue its research efforts to develop methods for,/
repository and waste packaging performance assessment in support df>*
the agency's high-level waste regulations and associated regulatory' guides.
6.
The severe accident research program must provide timely information in the Commission's decisionmaking process on severe accidents.
The senior management of user offices reviews and endorses the research program at two points in the planning process:
(1)LRRPand(2)budgetpreparatiorg User offices are asked to endorse the following in tire LRRP:
1.
That the regulatory needs are comprehensive and are accurately stated.
2.
That the priorities assigned each need are appropriate and the expected completion dates of supporting research are consistent with NRC needs.
3.
That the research product can reasonably be expected to provide the information needed to help resolve the associated regulatory need.
4.
That the level of expenditures for each program appears cost effec-tive relative to the research deliverables, resolution of the asso-ciated regulatory needs, and the inherent level of difficulty (experi-mental or analytical technique).
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Q The primary objective of NRC research is to support the regulatory process and contribute to improved reactor safety. The goal of the research planning pro-cess is to develop a program with a reasonable balance between near-term (those supporting current regulatory and licensing activities) and longer-term regula-tory needs.
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INTRODUCTION The mission of the Nuclear Regulatory Commission--regulation to ensure that civilian activities involving the use of nuclear materials and facilities are conducted in a manner consistent with protecting the public health and safety, the quality of the environment, and the national security--calls for the exer-cise of the regulatory functions of rulemaking, licensing review, and inspec-tion and enforcement.
In the process of carrying out its mission, the Commission makes policy deci-sions involving complex technical issues and varied and conflicting public attitudes.
The Commission must base these decisions on an accurate under-standing of the technical factors involved, and the NRC staff is responsible for ensuring that the Commission is fully informed.
In its responsibility for supporting the Commission's decisionmaking, the NRC staff must also maintain continuing awareness and understanding of public concerns and evolving understanding of issues that might signify a need for changes in the regulatory process.
In addition to maintaining a state-of-the-art analytical capability to ensure the technical accuracy of its assess-ments, the staff must use those new insights gained from operating experience as a basis for reassessing technical criteria with the goal of improving the regulatory process.
The staff is assisted in these areas by the research program of the Office of Nuclear Regulatory Research (RES).
The research program provides the technical basis for rulemaking and regulatory decisions to support licensing and inspection activities, to assess the feasi-bility and effectiveness of safety improvements, and to increase our understand-ing of phenomena for which analytical methods are needed in regulatory activities.
The major objective of the NRC research program is to provide an understanding of phenomenology and verified analytical methods to permit identification of important accident sequences and well-founded realistic (or best-estimate) analysis of their consequences. To this end, much of the research program consists of a mixture of experimental work and code development work aimed at understanding complex system transients.
Because the data points from large, complex, integral facilities tend to be few in number and of limited appli-cability, current and future research is based on experiments with a smaller scale than some earlier experiments to ensure cost effectiveness.
The data obtained will be used to validate codes for use in safety analyses.
Other objectives are to provide the methodology to make more effective use of prob-abilistic risk assessment in the regulatory process and to improve confidence in the data base for risk assessment.
This combination of experiments, code work, and risk analyses will produce thoroughly validated codes for use by licensing reviewers and will identify the areas in the regulatory process where improvements are needed.
Development of this Long-Range Research Plan (LRRP) is an important step in the process of ensuring that the Commission's research program is directed toward areas of importance to the regulatory program.
The LRRP is intended to assist ix
the Commission in establishing priorities to ensure effective utilization of limited resources.
It identifies broad regulatory issues and describes program-matic approaches for research to support the resolution of these issues over a 5 year period.
The LRRP is distributed broadly within the NRC for review and comment.
In addition, the final plan is distributed to DOE and to such industry groups as the Electric Power Research Institute (EPRI).
All recipients are invited to comment on the plan.
The Commission believes that DOE and the nuclear industry have a major responsibility to perform safety research to ensure that nuclear power plants and other nuclear facilities are designed and operated safely and reliably.
The distribution to DOE and the industry groups is intended to foster cooperation and coordination among NRC, DOE, and the nuclear industry to ensure that the appropriate level of effort is directed toward resolving safety issues and to prevent unnecessary duplication.
The LRRP is updated annually to reflect completed tasks, to identify new regulatory and research needs, and to incorpo-rate comments on the plan of the previous year.
The final selection of research programs is based on the Commission guidance; the needs submitted to RES by other NRC offices; the comments and technical insights from the Advisory Committee on Reactor Safeguards (ACRS), industry, the public, the national laboratories, and international organizations; and the availability of resources to ensure timely delivery of research results.
This LRRP is divided into eight chapters covering research programs such as operating reactor inspection, maintenance, and repair; equipment qualification; seismic research; reactor operations and risk; thermal-hydraulic transients; severe accidents; radiation protection and health effects; and waste management.
The proposed funding levels for the major research program areas described in the LRRP for Fiscal Years 1987-1991 are shown in Table 1.
Tat,le 2 provides the funding and staffing levels for FY 1987, and Table 3 is a summary sheet of possible regulation changes as a result of the research program described in this LRRP.
A correlation between needs and research products is provided in this LRRP, i.e.,
Research Product 1 refers to Need 1.
Dates
- are included after each statement of regulatory need and each research product. The date provided after the research product indicates when the research results should be available; the date provided after the statement of regulatory need indicates when the regulatory product should be available.
"FY (for fiscal year) has been omitted from the targeted dates provided in this LRRP.
It may be assumed, however, that the period of time indicated is the fiscal year.
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Table 1 LONG-RANGE RESEARCH PLAN FUNDING LEVELS
- FY 1987-FY 1991 FY 1987 FY 1988 FY 1989 FY 1990 FY 1991 Operating Reactor Inspection, Maintenance, and Repair
$ 24.9
$ 25.3
$ 26.5
$ 30.0
$ 31.0 Equipment Qualification 2.3 1.6 2.1 2.1 2.1 Seismic Research 11.7 11.5 11.2 11.2 11.2 Reactor Operations and Risk 8.2 9.8 9.0 9.0 9.0 Thermal-Hydraulic Transients 16.5 21.4 27.4 27.4 27.4 Severe Accidents 24.7 24.6 23.9 23.9 23.9 Radiation Protection and Health Effects 0.6 0.4 0.4 0.4 0.4 Waste Management 5.6 9.0
- 9. 0 9.0 9.0 TOTAL
$ 94.5
$103.6
$109.5
$109.5
$109.5
" Dollars in millions.
Xi
Table 2 FUNDING AND STAFFING FOR BASE YEAR FY 1987 (Dollars in Millions)
FY 1987 Staff 1.
OPERATING REACTOR INSPECTION, MAINTENANCE, AND REPAIR..............................................
24.9 22 1.1 Reactor Vessels..................................
7.8
- 1. 2 Steam Generators.................................
0.7 1.3 Piping...........................................
4.3 1.4 Electrical and Mechanical Components.............
8.9 1.5 Nondestructive Examination.......................
3.2 2.
EQUIPMENT QUALIFICATION...............................
- 2. 3 5
2.1 Qualification of Mechanical Equipment (Environmental)................................
2.3
- 2. 2 Dynamic Qualification of Equipment...............
0 3.
SEISMIC RESEARCH......................................
11.7 11 3.1 Seismic Hazard...................................
- 3. 2 Seismic Risk.....................................
3.3 Seismic Margins..................................
4.
REACTOR OPERATIONS AND RISK...........................
8.2 40 4.1 Risk Methodology and Data Development............
4.2 Operational Safety Reliability...................
4.3 Severe Accident Risk Assessment..................
5.
THERMAL-HYDRAULIC TRANSIENTS..........................
16.5 17 5.1 Separate-Effect Experiments and Model Development....................................
5.2 Integral Systems Experiments.....................
5.3 Code Assessment and Application..................
5.4 Plant Analyzer and Data Bank....................
5.5 Technical Integration Center.....................
6.
SEVERE ACCIDENTS......................................
24.7 29 6.1 Severe Accident Sequence Analysis................
6.2 In-Vessel Core-Melt Progression.................
6.3 Hydrogen Generation and Control..................
6.4 Fuel-Structure Interaction.......................
6.5 Containment Analysis.............................
xii
Table 2 (Continued)
FY 1987 Staff 6.6 Fission Product Release, Behavior, and Transport.
6.7 Containment Failure Mode.........................
6.8 Fission Product Contro1..........................
6.9 Severe Accident Risk and Risk-Reduction Analysis.
7.
RADIATION PROTECTION AND HEALTH EFFECTS...............
0.6 15 7.1 Metabolism and Internal Dosimetry................
0 7.2 Health Effects Standards Development.............
0 7.3 Occupational Radiation Protection................
0.6 8.
WASTE MANAGEMENT......................................
5.6 19 8.1 High-Level Waste.................................
3.0 8.2 Low-Level Waste..................................
2.6 TOTAL
$ 94.5 158 l
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Table 3
SUMMARY
SHEET LRRP INDICATIONS OF REGULATION CHANGES AS RESULT OF RESEARCH CONDUCTED IN 1987 AND BEYOND LRRP Targeted LRRP Section Subject of Research Date Page 1.1 Reactor Vessels - PTS-validated 1987 1-1 fracture analysis methodology 1.1 Reactor Vessels - PTS-fracture 1988 1-1 toughness and crack arrest tough-ness of irradiated vessel steel and weld metal 1.3
. Piping validated analysis metho-1989 1-7 dology for loading capacity of flawed and degraded piping; validation of leak-before-break concept; and data on true failure modes of cracked piping 1.3 Piping - evaluation of aging and 1989 1-8 degradation in LWR materials 2.1 Qualification of Mechanical Equipment 1987 2-1 (Environmental) - evaluation of proposed methods of qualifying equipment for design basis events 2.2 Dynamic Qualification of Equipment -
1987 2-3 evaluation of proposed methods of qualifying equipment for design basis events 3.1 Seismic Hazard - data concerning 1987 3-1 seismic zones in Eastern U.S.
3.1 Seismic Hazard - information bace 1990 3-2 for developing site specific spectra 3.1 Seismic Hazard - methods for handling 1990 3-2 uncertainties in assessing potential risk from seismic hazards xiv
Table 3 (Continued)
LRRP Targeted LRRP Section Subject of Research Date Page 6.3 Hydrogen Generation and Control - data 1987 6-6 from all areas of hydrogen research such as generation, ignition conditions, and mixing 8.2 Low-Level Waste - capability to assess 1989 8-7 alternatives to shallow-land burial of LLW xv
1.
OPERATING REACTOR INSPECTION, MAINTENANCE, AND REPAIR (PLANT AGING / LIFE EXTENSION)
Research is needed to study and understand time-related issues such as the mechanisms of aging and degradation, methods of examination and testing to determine the condition of components and structures, and interpretation of results of these tests for appropriate action.
This work will provide the bases by which the staff can assess with confidence industry test and examina-tion methods and results.
These assessments in turn will provide bases for licensing and relicensing decisions on whether operating plants continue to meet health and safety requirements in effect at the time of licensing and relicensing and subsequently imposed health and safety requirements.
1.1 Reactor Vessels This research applies to the structural integrity of pressure vessels espe-cially as affected by irradiation embrittlement and growth of postulated cracks in service.
1.1.1 Major Regulatory Needs and Their Justifications 1.
Validated methodology for fracture analysis of reactor vessels under acci-dent conditions, to provide a basis for the development of licensing cri-teria and a regulation and to assist the staff in assessing submittals implementing the actions taken to resolve unresolved safety issue (USI)
A-49 on pressurized thermal shock (1987).t Justification:
The validated fracture analysis methodology is needed to enable the staff to independently evaluate vendor and utility submittals concerning the ability of a reactor vessel to safely withstand conditions imposed during accidents.
Note that Section III of the American Society of Mechanical Engineers (ASME) Code is fundamentally a design code, and certain aspects of aging are beyond its scope; thus, safety criteria for operating vessels (which are subject to aging degradation) must be developed by the NRC.
2.
Data base on fracture toughness and crack arrest toughness of irradiated vessel steel and weld metal, to provide a basis for the development of safety evaluation criteria, an amendment to the regulations, a revision to Regula-tory Guide 1.99, and recommendations for a possible update of Section XI of the ASME Code and to assist the staff in assessing submittals implement-ing the actions taken to resolve USIs A-11 and A-49 (1988).t Justification: Generic and specific data are required for irradiation effects on the fracture initiation toughness and crack arrest toughness of vessel steels and weld metal so that safety decisions on vessel integrity can be made.
Otherwise, the staff would not know how much a vessel steel had degraded or what was an acceptable level to ensure continued safety.
tA modification of the regulations may result from the research.
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3.
Experimentally validated surveillance methodology to allow accurate pre-dictions of neutron fluence and radiation effects so as to provide bases for recommendations for updating American Society for Testing and Materials (ASTM) standards that are or will be endorsed by regulatory guides (1987).
Justification:
Neutron flux can be measured with reasonable accuracy in experimental facilities.
In these test reactor facilities, irradiated Charpy V notch specimens are tested to establish the empirical relation-ships between fluence and the reduction of the Charpy specimen's fracture toughness.
In an operating reactor, the core flux leakage calculations form the basis for our predictions of the fluence impinging on and through the vessel wall during the lifetime of the plant.
These predictions are validated periodically by comparison with dosimetry from the vessel's surveillance capsules and then using the developed relationships to estab-lish the fracture coughness.
Research has shown that significant errors exist in the calculational methods used to predict fluence and radiatio'n effects.
Further research will reduce the error band in these calculation methodologies and hence significantly improve the ability to accurately assess the structural integrity of these structures at any period in their expected life during both normal and postulated accident conditions.
4.
Data base on environmentally assisted fatigue crack growth rate in vessel steels and welds, to be the basis for safety evaluation criteria for aging plants and for recommendations for updating Section XI of the ASME Code (1988).
4 Justification:
Knowledge of the rate of environmentally assisted fatigue crack growth in nozzles, piping, and vessels is necessary to decide if cracks or flaws discovered during inspections can grow to critical size in subsequent operation (and thus must be removed) or if they can be allowed to remain as benign imperfections with no potential impact on the safety' of the primary system during normal operations or accidents.
5.
Validated methodology for annealing to recover material fracture toughness properties of irradiated reactor vessels, to develop licensing criteria and recommendations for updating ASTM standards and Section XI of the ASME Code (1989).
Justification: The validation of annealing methodology for recovery of properties degraded due to vessel irradiation is needed so that safety decisions on vessel integrity can be made based on actual engineering test data and with a minimum of risk of impacting continued safe operation.
i 6.
Data base on parameters and light-water reactor (LWR) environmental conditions that result in stress corrosion cracking of ferritic steels used for pressure boundaries, to provide a basis for developing regulatory criteria and assisting staff in assessing failures in ferritic components (1989).
Justification:
Generic and specific data are required for environmental effects on the initiation and growth of stress corrosion cracking in pressure boundary ferritic materials so that decisions on the integrity of pressure boundary components can be made and design assumptions and design life may be verified.
1-2 2
n
7.
Crit.eria and corresponding data base for evaluating applications for i
plant life extension of reactor vessels and associated components (1990-1991).
Justification:
It now seems quite clear that many utilities will elect to renew nuclear plant licenses when the original operating licenses (0Ls) begin to expire in the early 1990s, rather than scrap these investments costing nearly a billion dollars.
The NRC staff must soon be in a position to define critical issues for discussion by applicants to establish their cases for OL renewals.
Furthermore, the NRC staff must have a solid background of experimental data to use as a basis for application review and for NRC decisions.
1.1.2 Research Program Description The ability of the NRC licensing staff to make decisions concerning the present and continuing safety of reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data.
This program is to provide both the analytical methods and the experimental data needed.
Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operating conditions.
To do this, not only must the methods be developed but they must be experimentally validated.
Further, the materials data necessary for input to these analytical methods must also be developed.
Thus, in addition to methods development and large-scale experimental verification, this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness char-acteristics of the material behavior in pressure vessels in both the unirradiated and irradiated conditions.
Another significant role of this program is to determine the ways and the extent to which the LWR environment (particularly radiation) changes and degrades the pressure vessel materials during their operational life.
- Thus, elements of this program deal with the determination of the sensitivity of the pressure vessel's steel to fracture toughness degradation as a function of irradiation exposure and with methods such as thermal annealing required to regain this toughness parameter.
Also part of this program are studies to improve and standardize dosimetry, damage correlation, and the associated reactor analysis procedures used for predicting the integrated effects of neutron exposure to these steels.
The strategy for the research in this element is (1) to develop experimentally verified fracture mechanics analysis techniques that can be rapidly brought to bear in resolving licensing issues dealing with the assurance of reactor pressure vessel integrity during normal operation and postulated upset or accident conditions and (2) to establish statistically acceptable material data bases to be used in conjunction with the presently accepted and newly developed analytic techniques.
1-3
The research effort is divided into three phases:
(1) relatively short-term, high priority programs to develop improved methodologies and sufficient supporting data to be used in establishing generally acceptable and defensible regulatory positions on critical issues such as pressurized thermal shock, structural performance of low upper-shelf energy materials, and irradiation embrittlement rates of presently operating reactor pressure vessels; (2) a longer-term effort to establish procedures for degraded material fracture toughness recovery and the revision or replacement of existing standards, codes, and criteria that deal with the fabrication and expected aging effects of reactor pressure vessels; and (3) a new facet of the longer-term effort devoted to establishing criteria and data bases upon which to base decisions for plant life extension.
Incorporated into this research program will be work done by outside groups (Electric Power Research Institute (EPRI), several European governments and the Japanese government, and the efforts of European technical commmunity groups such as the Program for Inspection of Steel Compo-nents (PISC)), which is anticipated to approximately equal the NRC research effort in costs.
The short-term, high priority effort will be completed by 1987, the longer, confirmatory research reaching completion by 1989, and the plant life extension research making initial milestones in 1990-1991.
The major research products will be:
1.
a.
Large-scale verification of unified fracture mechanics methodology by completion of the Wide-Plate Test Series (1988),
b.
Completion of development of a "visco plastic" mechanism model to define rate effects on propagation and arrest of rapidly running crack (1988).
c.
In cooperation with industry, complete development of statistical data base defining defect distribution in as-fabricated reactor pressure vessels (1989).
2.
a.
Establishment of basis for NRC position on ASTM standard on crack arrest testing specimen (1987).
b.
Irradiated specimen test completed to validate ASME Section XI fracture toughness curves (1987).
c.
Technical bases for revising 10 CFR Part 50 governing reactor fracture toughness requirements under both normal and accident operating conditions (1988).
d.
Development of crack arrest data base for irradiated steels and weldments for first generation and present practice reactor pressure vessels (1989).
e.
Review and evaluation of industry-developed fracture toughness data base for new, high-strength materials for next generation reactor pressure vessels (1990).
1-4
3.
a.
Establishment and maintenance of well-defined reproducible benchmark experiments, methodologies, and data base to predict damage fluence parameters for pressurized thermal shock (PTS) embrittlement extra-polation ant. reactor vessel embrittlement predictions (1987).
(Also applies to Needs 1, 2, and 7.)
b.
Development of basis for drop in upper-shelf energy toughness criterion (1987).
c.
Complete development of irradiation damage mechanism model to allow accurate transition temperature shift predictions beyond empirically developed trend curves (1988).
(Also applies to Needs 2 and 7.)
d.
Extension of interpretation of data bases and irradiation damage mechanism model to allow prediction of effects of irradiation damage to post-heat-treat annealed reactor pressure vessels (1990).
(Also applies to Need 5.)
4.
a.
Revision of fatigue curves for ferritic materials in ASME Section III (1988).
b.
Final revisions recommendet for the environmentally assisted fatigue curves in ASME Section XI (1988).
5.
a.
Development of nethodology (engineering design and instrumentation requirements) for guidance on cooperating with industry in a full-scale vessel annealing experiment (1988).
b.
In cooperation with industry, completion of construction of full-scale vessel annealing equipment (1989).
c.
In cooperation with industry, completion and experimental validation of methodology for recovering fracture toughness properties by in situ annealing (1990).
6.
a.
Initial findings on environmental effects on stress corrosion cracking of ferritic materials (1987).
b.
Licensing criteria proposed for establishing limits on environmental variables to control cracking on pressure boundary components (1988).
7.
a.
Criteria on reactor vessel and associated component material degradation and structural response behavior considerations defining issues that must be adequately addressed for plant life extension (1989-1990).
b.
Data base on reactor vessel and associated component material property changes and structural response to normal operations and accident conditions to be expected over the span of the new OLs to be granted for plant life extension (1989-1991).
1-5
1.2 Steam Generators The research discussed below deals with corrosion, cracking, and degradation of steam generator tubing during service and integrity of tubing as degraded by the water and stress environment during normal operation and upset conditions.
- 1. 2.1 Najor Regulatory Needs and Their Justifications 1.
Validated data on integrity of tubing having cracks, dents, wastage, and other forms of degradation, to be the basis for licensing criteria and revisions to regulatory guides (1987).
Justification:
Steam generator tubes have been and are degrading in the form of cracking, pitting, intergranular corrosion, and general wastage.
The staff must know the potential remaining integrity in tubes having various degrees and types of degradation, cracks, etc., for requirements on tube plugging and additional inspection of tubes.
If inspection
" indications" translate to potential cracking and leakage, plugging or augmented inspections would be required; if indications are benign, the plant can be returned to service.
2.
Correlation of nondestructive examination (NDE) signals with tube integ-rity, to be the basis for licensing criteria and revisions to regulatory guides (1987).
Justification:
The only way to predict tube integrity is from knowledge of the signal taken from inspection.
Thus, signal evaluation errors must be reduced, and signals must be carefully correlated to defect type and size, as well as to measurements of tube burst and collapse strength.
1.2.2 Research Program Description Research on steam generators at NRC is focused on the Steam Generator Group Project at Richland, Washington, where four cosponsors from the United States and abroad have joined with NRC to conduct the program described below.
The program aims to develop validated models, based on experimental data, for pre-dicting margins-to-failure under burst and collapse pressures of steam generator tubing found to be service degraded by eddy current inservice inspection (ISI).
This is to be accomplished by using an out-of-service degraded steam generator as a test bed for a confirmatory research program that includes NDE improvement and validation; optimization of ISI procedures, sampling plan, and inspection period; validation of tube integrity predictive models; and optimization of tube plugging criteria.
EPRI has under way some steam generator reliability work, a large amount of which may be validated in the NRC steam generator project through EPRI's parti-cipation in that project.
The project has become a focus for international efforts in steam generator research as France, Italy, and Japan, as well as EPRI, have joined the work through financial contributions.
Specifically, service-degraded tubes in a retired-from-service steam generator will be used to compare and validate different and advanced NDE methods that will show the best methods for detecting and characterizing flaws, to remove these tubes and precisely characterize the type and extent of cracking or other 1-6
degradation, and to subject the tubes to pressure in burst or collapse mode to establish the residual strength in the tubes.
In this manner, a correlation can be developed between the flaw signal, as detected and evaluated during inservice inspection, and the tube integrity.
Thus, the licensing criteria for tube inspection plans and tube plugging can be validated or modified, using the research results, to reflect levels of integrity needed for safety.
Of particular value is the ability to use service-degraded tubes with flaw and degradation characteristics that have been carefully documented to validate advanced eddy current NDE methods as well as models for predicting stress cor-rosion cracking in tubing.
(NDE research for steam generators is also discussed in Section 1.5.)
The major research products will be:
1.
a.
Burst and leak rate testing of tubes removed from generator (1987).
(Also applies to Need 2.)
b.
Validation of current and advanced NDE results through examination of removed tubes (1987).
(Also applies to Need 2.)
2.
Correlation of remaining tube integrity from burst and leak tests with NDE results to validate regulatory guide ISI plans and tube plugging criteria (1987).
1.3 Piping This research applies to the structural integrity of piping degraded during service by the water, stress, and temperature environment.
This degradation is in the form of stress corrosion cracking, fatigue and cyclic crack growth, and toughness loss because of long-time aging at temperature.
Evaluation of the factors causing cracking and of proposed fixes is included.
Pipe rupture investigations are also a part of this research program.
- 1. 3.1 Major Regulatory Needs and Their Justifications 1.
Experimentally validated analysis methodology for the loading capacity of flawed and degraded piping during normal operation, accidents, and earth-quakes; evaluation of leak detection systems for use inside and outside containment; evaluation of leak rate estimation models; and data on the true failure modes of cracked piping, to provide the basis for new or modified regulatory guides, regulations, and standard review plans (1989).t Justification:
Decisions are regularly needed on the safety of pipes and welds containing flaws or cracks discovered during inservice inspections.
Even if a cracked pipe could withstand normal operating loads, it might not be able to withstand the loads from all postulated accidents and earth-quakes.
These concerns have been heightened by the occurrence of stress corrosion cracking in large-diameter BWR piping.
Recent acceptance of leak before break to eliminate dynamic effects for primary loop piping in PWRs has heightened interest in accepting leak before break for the balance of plant and raised questions regarding the extension of the concept to TA modification of the regulations may result from the research.
1-7
r other areas such as emergency core cooling system (ECCS) sizing, contain-ment design, and equipment qualification.
The consideration of these areas requires the development of a replacement to the existing pipe rupture criteria.
2.
Data base on crack growth rate in piping steel and welds, to be used in developing licensing criteria and recommendations for updating Section XI of the ASME Code (1988).
Justification:
Knowledge of rate of growth of cracks under operating environment and loading conditions is necessary to decide if cracks or flaws discovered during inspections can grow to critical size in sub-sequent operation (and thus must be removed) or if they can be allowed to remain as benign imperfections with no potential impact on the safety of piping during normal operations or accidents.
3.
Data base for evaluating toughness loss in existing cast duplex (austenite I
and ferrite) stainless steel components from long-term aging at reactor operating temperatures and for making recommendations to limit long-term i
aging ~in new components, to be the basis Tor developing a regulatory guide and revising the standard review plan (1988).
Justification:
A certain level of toughness is required to ensure safety in piping and other primary system components, especially to resist failure 3
i if flaws should develop in service and under accident loading.
Long-term j
time-at-temperature can cause precipitation of other phases in the ferrite of the duplex stainless steels and at grain boundaries resulting in a reduced toughness of the original material.
The time-temperature-material conditions under which this occurs and the degree to which it occurs in 4
l service must be known so that licensing decisions are made in full know-ledge of the future strength and toughness condition of piping and other primary system components such as pump casing.
4.
Evaluation of aging and environmental degradation in LWR materials, includ-ing the effects of temperature, irradiation, and environment, to be the Fasis for changes in regulations dealing with operations and maintenance of nuclear potter plants (1989).t Justification:
Recent reports have highlighted the possible detrimental effects of hydrogen on stress corrosion cracking of Inconel 600, of irra-diation on intergranular stress corrosion cracking (IGSCC) of annealed 304 SS, and of thermal aging on toughness loss of ferritic steels.
These phenomena need to be evaluated for their effect on components and mate-rials of interest in the primary system.
The degree ~of degradation as a function of time caused by these phenomena needs to be established for proper safety analyses.
I 5.
Evaluation of the types and sources of fluid system leakage, the expected i
leakage rates, and the rate of increase of these leakage rates (1990).
Justification:
Public comments on the limited scope modification of General Design Criterion 4 raised seven issues considered by the Commis-i sion.
One of those issues was the extension of the rule to relax pipe rupture requirements for containment design, ECCS performance, and environ-4 mental qualification of electrical and mechanical equipment. The Commis-sion response, in part, noted that further studies must be conducted to i
tA modification of the regulations may result from the research.
1-8 i
f n---
develop suitable replacement criteria for the PWR primary coolant loop double-ended pipe rupture if this accident is no longer required for containment design, emergency core cooling, or environmental qualifica-tion. This research activity will contribute to the technical basis for defining such a replacement.
6.
Evaluation of small-break loss-of-coolant-accident (LOCA) probabilities and associated leak rates, to provide input to the formation of a technical basis for suggesting a replacement to the existing pipe break criteria (1989).
Justification:
Extending the leak-before-break concept to areas such as equipment qualification requires definition of a replacement to the exist-ing pipe break criteria.
Definition of the replacement criteria must consider the expected types of fluid system leakages, which includes leaks from cracked piping.
The experimental programs being conducted will pro-vide a deterministic methodology that can be used to define expected leak-age from bolted flanges, flange joints, valves, pumps, etc., as well as from pipe cracks and small breaks.
However, credible pipe cracks and small breaks must be considered.
The probabilistic methodologies previously developed provide an ideal tool for this purpose.
1.3.2 Research Program Description The principal long-term objectives of the piping research program are to deter-mine the applicability of the leak-before-break concept in LWR piping systems and to provide the capability to evaluate potential fabrication and operating improvements directed at eliminating pipe cracking.
The program for evaluating leak before break in LWR piping systems is a multifaceted effort that will integrate research in the areas of piping degradation modes, piping fracture mechanics, nondestructive examination, leak types and sources, leak rates, and leak detection.
A major program to validate elastic plastic fracture mechanics analyses and to develop a material properties data base for piping was initiated in 1983. A complementary program is being developed to address seismic and dynamic loading effects and leak rate estimations over a broad range of leak l
rates.
This research effort is being jointly funded by a consortium of inter-l national organizations interested in developing an international consensus on leak-before-break technology.
Initially these programs are addressing materials that exist in operating plants.
The fracture properties of new and replacement I
materials will be evaluated during later phases of the programs.
Ongoing and planned programs will provide information on NDE techniques, leak rates, and i
leak detection systems.
The integration of results from these programs will be I
a continuing effort, culminating in a position on the overall acceptability of leak before break as a function of the piping system, material of fabrication, i
and other pertinent factors.
Appropriate develcpment of regulatory guides,
{
modifications to the standard review plan, and rulemaking will then be pursued.
Research programs directed at environmentally assisted crack growth, effects of welding on sensitization and susceptibility to cracking, and aging effects in piping will provide the necessary basis for evaluating the acceptability of fixes proposed by the industry to eliminate or reduce the frequency of pipe cracking and to eliminate the degree of age-related degradation in piping materials.
The data to be generated will be applicable to evaluating improved fabrication and repair procedures, proposed new materials, and changes in operating environnent.
1-9
Elements of the research program are directed toward providing timely informa-tion to assist in the development of licensing criteria regarding stress corro-sion cracking in BWR piping.
These elements include determination of the reli-ability of detection and sizing of stress corrosion cracks, identification of i
improved NDE techniques, development of data on stress corrosion crack growth rate, validation of fracture mechanics analyses for evaluating stress corrosion cracks, and evaluation of short-term and long-term fixes.
The short-term and long-term fixes of interest include weld overlays, induction heating stress improvement, last pass heat sink welding, oxygen control, and piping replacement.
The objective of piping-related investigations is to support the implementation of the leak-before-break hypothesis in the revision of General Design Criterion 4 of Appendix A to 10 CFR Part 50 and to contribute to the technology base needed to consider a replacement to the existing pipe break criteria.
All piping research will have an additional focus-providing input to establish criteria and data bases upon which to base decisions on plant life extension.
The major research products will be:
1.
a.
Evaluation of pipe cracking predictive models, proposed fixes, and weld overlay criteria (1987).
(Also applies to Needs 2 and 3.)
b.
Final data development on toughness of aged cast stainless steels, including austenitic and other nickel alloy weld material deposited using a flux process such as SMAW (Shielded Metal Arc Welds) and SAW (Submerged Arc Welds), for use in leak-before-break study (1989).
(Also applies to Need 4.)
c.
Final conclusions and recommendations regarding the effectiveness of short-term fixes for pipe cracking (1987).
(Also applies to Needs 2 and 3.)
d.
Experimental validation for elastic plastic fracture mechanics analyses, including effects of seismic and dynamic loadings and prototypical geometries (1989).
Computerized data base on piping materials fracture toughness and e.
crack growth rates transmitted to NRR for use in licensing evalua-tions (1987).
l f.
Evaluation of effectiveness of long-term fixes for pipe cracking (1987-1988).
(Also applies to Need 2.)
g.
Technical basis for licensing decision on acceptance of leak before break in LWR piping systems (1988).
(Also applies to Need 7.)
h.
Validation of ductile fracture mechanics analyses for complex piping geometries and components (1989).
i.
Evaluation of detection sensitivity, repeatability, and reliability of potential leak detection devices for use inside and outside containment (1989).
1-10
2.
Licensing criteria proposed for establishing limits on environmental vari-ables to control pipe cracking in LWR piping systems (1987).
3.
Licensing criteria proposed for prevention and control of toughness degra-dation due to aging in LWR cast stainless steel piping materials (1988).
(Also applies to Need 1.)
4.
a.
Evaluation of the effect of different system characteristics coupled with hydrogen on stress corrosion cracking (SCC) of BWR piping materials (1986-1987), of thermal aging on ferritic materials, and of irradia-tion on SCC of internal stainless steel materials (1989).
b.
Recommendations on degree of degradation induced by hydrogen, aging, and radiation exposure of LWR materials and on its significance to structural integrity and safety (1989).
5.
Experimental evaluation of potential leakage types and sources, leakage rates that can be expected from these sources, and potential rate of increase of these leakage rates that can be used in considering revision to containment design criteria, sizing of ECCS system, and establishing suitable environmental profiles for equipment qualification (1990).
6.
Probabilistic evaluation of small-break LOCAs coupled with deterministic methodologies that can contribute to licensing decisions on a replacement to the existing pipe break criteria (1989).
1.4 Electrical and Mechanical Components (Nuclear Plant Aging Research - NPAR)
This generic nuclear plant aging research applies principally to the time-related degradation of electrical and mechanical components during service and the potential impacts of degradation of plant systems involving these ccmpo-nents upon public safety.
The major goals of the program are to (1) identify aging and service wear effects associated with electrical and mechanical compo-nents and systems that, if unt.hecked, could impair plant safety, (2) identify methods of inspection, surveillance, and condition monitoring of electrical and mechanical components and systems that will be effective in detecting signifi-cant aging and service wear effects prior to loss of safety function so that proper maintenance and timely repair or replacement can be implemented, and (3) identify and recommend acceptable maintenance practices that can be undertaken to mitigate the effects of aging and to diminish the rate and extent of degrada-tion caused by aging and service wear.
The technical knowledge gained from these tasks will be translated into practical application guidelines to monitor and mitigate the aging of components and structures in nuclear power plants.
The technical knowledge will also be translated into recommendations for standards and guides concerned with monitoring equipment degradation and the prediction, prevention, and mitigation of equipment failures that can adversely affect public health and safety.
1-11 i
i The general nuclear plant aging research program
- will be coordinated with other aging research under way in NRC, in other United States Government agencies, in industry, and in other countries.
Eventually, plants will age sufficiently th'at utilities will choose to decom-mission them.
The goals of the decommissioning program are to supply the infor-mation on dose, labor requirements, techniques, waste amount, and costs for use in preparing rule amendments, regulatory guides, and an NRC staff evaluation of funding assurance plans and end-of-life decommissioning plans submitted by licensees.
1.4.1 Major Regulatory Needs and Their Justifications 1.
Assurance that previously unaccounted for aging and service wear effects that could have a significant impact on safety over the life of a nuclear power plant are identified so that modifications to regulatory requirements to mitigate such effects may be developed on a timely basis (1987-1990).
Justification:
Although aging effects have previously been recognized to be potentially important to nuclear plant safety, only a limited number of e
Institute of Electrical and Electronics Engineers (IEEE) and ASME standards provide guidance on how to account for aging of electrical and mechanical components.
In the case of the IEEE standards, emphasis has been placed on pre-aging (primarily accelerated aging based on the Arrhenius model) prior to qualification testing.
However, it is generally recognized that all aging and service wear effects cannot be modeled within the context of the Arrhenius theory.
Both IEEE standards and the ASME operations and maintenance standards have recommended needs for equipment surveillance, degradation monitoring, and maintenance.
However, the guidance on surveil-lance and maintenance in these standards is very general in nature, and specific indicators of incipient aging-related defects prior to catastrophic failure modes are not provided.
Also, standards have been developed to date for only a limited number of component types.
Taking into considera-tion the multiplicity of equipment types and the variety of instances of aging-related equipment failures or precursors of such failures reported over the years in licensee event reports (LERs), maintenance records, and inspection reports, it is not clear that the national consensus standards have provided adequate guidance on how to account for the aging effects that could degrade plant safety over the expected 40 year life of a typical commercial nuclear plant.
A systematic evaluation is needed to identify potentially significant aging effects, i.e., those that could cause an increase in frequency or severity of plant transients or an unacceptable degradation in the capability of safety equipment to withstand or mitigate design basis events.
Such an evaluation should include consideration of the severity of aging processes to equipment and components, the impact of equipment degradation on safety system performance, and the overall potential impact on risk to the public.
This evaluation would provide a basis for judgment that the relative importance to safety of the identified aging effect has been appropriately characterized and would guide the development of regulatory requirements.
- See NUREG-1144, " Nuclear Plant Aging Research Program Plan."
1-12
.. ~..
4 i
2-2.
Criteria to be used as the technical basis for evaluation of industry surveillance testing and monitoring, maintenance, and replacement programs to determine whether those programs adequately mitigate aging and service wear effects that could have a significant impact on plant safety (1988-1991).
Justification:
National consensus standards include limited guidance i
regarding surveillance, maintenance, and inservice inspection to account for aging; nuclear plant technical specifications include requirements for
}
periodic surveillance and testing; and utilities have instituted mainte-E nance and replacement programs. however, no regulatory criteria have been develcped for evaluating these programs to determine whether significant aging effects can be adequately mitigated. Also, there is general recogni-tion that artificial pre-aging techniques based on the Arrhenius theory do not realistically simulate actual aging and service wear processes and degradation for all types of components and systems.
Although artificial pre-aging is currently applicable primarily to electrical equipment to be qualified for harsh environments, consideration is being given to whether mechanical equipment (or at least the nonmetallic materials contained in such equipment) should be pre-aged prior to testing for harsh environments i
and to whether equipment located in mild environments should be aged prior t
to seismic and dynamic qualification.
Criteria are needed to (1) evaluate the surveillance, maintenance, and replacement programs instituted by nuclear utilities to determine whether these programs will adequately prevent significant impairment of safety function; (2) evaluate which, if any, surveillance and testing intervals in plant technical specifications should be modified as a function of plant age to account for or to reduce potential aging and service wear 4
i effects; and (3) evaluate surveillance monitoring programs developed to supplement and perhaps partially replace artificial pre-aging of equipment i
prior to qualification testing.
(As noted above, current requirements i
differ depending on whether the equipment is mechanical or electrical and whether the environment is harsh or mild.)
1' In all cases, the evaluation criteria should be limited to equipment i
determined likely to be vulnerable to aging effects that could cause significant impact on plant safety.
In addition, for surveillance programs the criteria should be based on indicators of aging or service wear degradation that can be monitored at reasonable cost and with the minimum possible accumulation of occupational exposure to radiation.
Proper condition monitoring techniques should effectively indicate the j
approach of a level of degradation that would render the equipment incapable of performing its safety function during design basis accidents.
i l
For equipment for which surveillance is impractical as a means of deter-i mining with confidence the approach to an unacceptable level of degrada-l tion, replacement and maintenance schedules may have to be based on the concept of " predicted service life." Although theoretical bases for predicting effective service lifetimes have been and are being developed, considerable uncertainties exist in the resultant predicted lifetimes because of the statistical inadequacies in the data base.
Criteria are needed for evaluating such predictions, and such criteria should be based on an adequate technical data base, including previous experience from operating plants and analysis of aged equipment.
L 1
1-13 5
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--~_.._.._. ~,,_ m-.
,_.m
..,_.,-_._m,.
.._,..,,-.,._-,,_.,,m,
.~.
(It is recognized that ASME Code Section XI Rules for Inservice Inspection of Nuclear Power Plant Components cover safety-related mechanical compo-nents.
Therefore, research in this area, when appropriate, will be conformatory in nature.)
3.
Improved predictions of long-term deterioration of sealer materials and l
of limits on hostile environmental exposures (1987).
Justification:
This information will help address the evaluation of containment integrity under severe accident environment.
4.
Basis for determining adequacy of funding plans and end-of-life decommis-sioning plans submitted by utilities through collection of actual decommis-sioning data (e.g., Humboldt Bay, Shippingport, TMI)'(1990).
Justification:
Decommissioning rules and regulatory guides needed to implement the rules are being developed.
The NRC needs current and actual information to evaluate licensee decommissioning activities and plans.
i 5.
Basis for establishing financial requirements for decommissioning and 1
the establishment of guidance on steps that should be taken to facilitate decommissioning during design, operation, and actual decommissioning (1990).
Justification:
Final decommissioning rules and regulatory guides neces-sary for rule implementation are being developed.
Information from licensees on financial assurance and facilitation is required by these rules, and the NRC needs current information to adequately evaluate licensee decommissioning activities and plans in these areas.
1.4.2 Research Program Description The strategy for this research (see NUREG-1144) is to develop a technical data base to predict in a timely manner the onset of significant component aging and service wear phenomena that can adversely affect public health and safety and to develop guidelines and criteria for surveillance testing and degradation monitoring, maintenance, and replacement programs to mitigate aging and service wear effects in electrical and mechanical components important to ensure plant
- safety, i
The initial phase of this research effort will provide input to the data base for the aging assessment of components and systems currently under way.
This study will identify significant component / environment aging mechanisms that can lead to the inoperability of vital electrical and mechanical components.
The study will also yield recommendations, including priorities and schedules, for further specific research.
Specifically, the approach and key activities to i
address aging assessment of critical electrical and mechanical components l
include the following:
o Selection of equipment to be studied.
l System / aging / risk-oriented evaluations and setting of priorities for components and structures to be investigated.
Acquisition of pertinent knowledge of experts through workshops, questionnaires, and licensing reviews.
1 1
1-14
~ - - - -
Review and analysis of existing aging data from operating experience, including LERs, reported occurrences, and evaluation of maintenance, refurbishment, and replacement programs that contend with the degrada-tion of components and structures.
Review of applicable codes, standards, and guides.
Development of technical basis for comprehensive assessment of aging of o
nuclear power plant components and structures.
l Review ano analysis of equipment specifications, designs, and operating parameters.
I Postservice examination and laboratory testing of aged equipment from decommissioned (e.g., Shippingport) and operating facilities to deter-mine aging mechanisms and aging-related failure modes.
Review and application of past and ongoing research on aging of materials and components.
Cooperative collection of data by monitoring equipment on site at one or more operating reactor facilities.
o Development of application guidelines and criteria for detection and mitigation of functional degradation of electrical and mechanical components with high-risk factors before major safety problems develop.
Evaluation of existing degradation monitoring techniques that would be effective in identifying functional degradation and the remaining functional capability of components.
Determination of practical, cost-effective indicators of functional capability.
Analysis of applicable codes, standards, guides, and industry practice.
Recommendations for advanced methods of condition monitoring.
Risk / cost / benefit analysis for practical and cost effective degrada-tion monitoring techniques.
Cooperative degradation monitoring and surveillance programs to moni-tor critical equipment on site at one or more operating reactor facil-ities to confirm feasibility of techniques.
Acquisition of data from decommissioning Humboldt Bay, Shippingport, and o
German reactors on dose rates, labor requirements, techniques, waste disposal, waste shipping, and costs and determination of the amounts and distribution of induced radioisotopes in samples of Shippingport components.
1-15
o Acquisition of information on current techniques and costs of decom-missioning nuclear facilities and on occupational doses associated therewith and information on methods that facilitate decommissioning by reducing doses and waste volumes.
The major research products will be:
1.
a.
Setting of priorities and selection of electrical and mechanical components important to safety and susceptible to functional degrada-tion due to aging (1987-1988).
b.
Comprehensive aging assessment of selected plant components, including postservice examinations and failure mode analyses (including assess-ment of significance of aging as a factor in capability to withstand seismic and dynamic stresses) (1987-1990).
2.
a.
Identification of practical techniques for monitoring equipment for aging and service wear effects (1987-1990).
b.
Assessment of effectiveness of surveillance monitoring in supplement-ing artificial pre-aging (prior to qualification testing) for selected electrical equipment (1988) and mechanical equipment (1990).
c.
Assessment of necessity to modify surveillance and testing intervals of selected systems in plant technical specifications to account for age (1988-1990).
d.
Assessment of methodologies for predicting service lifetime of equip-ment (1988-1991).
e.
Evaluation criteria for surveillance, maintenance, and replacement programs for selected components (1989-1990).
3.
Information on long-term material deterioration of sealer materials and limits on hostile environmental exposures (1987).
l l
4.
Information on actual decommissionings of nuclear reactors to be used to update data base for reviewing decommissioning activities and plans (1990).
5.
Current information on safety, costs, and facilitation of decommissioning, i
to be used in data base for reviewing funding assurance and decommission-ing plans (1991).
- 1. 5 Nondestructive Examination 1
i This research applies to the validation of reliable, reproducible NDE techniques for detection and characterization of cracks and flaws, etc., for pressure vessels, piping, and steam generator tubing as well as the associated inter-pretation and analysis for decisionmaking.
Assuming the current industry effort in this area will continue, the NRC research role is expected to significantly decrease in the 1987-1988 period.
j 1-16 t
I 1
1.5.1 Major Regulatory Needs and Their Justificationj 1.
Documentation and upgrading of the reliability and reproducibility of ultrasonic and eddy current inspection methods during preservice and inservice inspections for detection and characterization (sizing, orienta-tion, etc.) of flaws, cracks, and other defects.
The planned and corpleted research work formed the basis for resolving and dropping USI A-14,
" Nondestructive Examination," and will provide the basis for revising regulatory guides and for recommendations for updating Section XI of the ASME Code (1988).
Justification: Methods currently in use for flaw detection and character-ization are not necessarily always consistent, reproducible, or interpret-able.
Nevertheless, preservice and inservice inspections are counted upon to find and characterize flaws in reactor components. For safety evalua-tions such as for pressurized thermal shock (PTS), it is very important to know if the very small flaws capable of crack initiation under PTS accident conditions are present or not.
Thus, the methods currently in use must be quantified with respect to their reliability.
2.
Criteria and validation for use of acoustic emission for continuous monitoring for cracking in vessels and piping, to provide the basis for licensing criteria, amendments to technical specifications, and recommen-dations for changes to the ASME Code (1987-1988).
Justification:
Locations exist in plants where conventional inspection techniques are inadequate for proper examinations for flaws.
Thus, alter-native techniques are very useful.
One such technique is acoustic emission.
Here, a growing crack will produce an acoustic signal that can be monitored to produce warning, or a leak will also cause an acoustic signal that can be detected. Although such methods are desirable, no criteria exist for acceptance by NRC or for operation of the techniques in service, nor are the parameters and their appropriate useful ranges listed and justified.
3.
Development and validation of NDE techniques for conducting accurate and reliable "end-of-licensing period" nondestructive baseline examinations (for consideration of life extension) of reactors to establish the condi-tion of primary system components with respect to (1) any flaws or cracks initiated during prior service and (2) the material toughness and other mechanical properties after initial service (1991).
Justification:
To make licensing decisions related to extension of service for nuclear power reactors beyond their first licensing period, evaluations of the integrity of the primary system reactor components and their fitness for continued use must be performed to ensure continued safety.
To perform these evaluations, the condition of the components with respect to flaws initiated during prior service and the extent of any degradation in material toughness and other mechanical properties must be accurately known.
Nondestructive techniques for these assessments need to be developed and validated.
1.5.2 Research Program Description The strategy for this research is to establish the reliability of current tech-niques and procedures for NDE, especially those embodied in the ASME Ccde, and to validate improved or advanced techniques and procedures so that better accuracy of inspection can result and so that less conservatism need be applied in licensing decisions wherein flaw size and location are issues.
The NRC 1-17
research program is well coordinated with the major efforts under way in the United States, especially at EPRI, and also with major overseas efforts, especially the PISC efforts of the Organisation for Economic Co-operation and Development (0ECO) in Paris, France.
i The research approach is twofold:
(1) A series of test plates and pipes are prepared with known flaws for round-robin detection and characterization trials l
from which conclusions can be drawn about the reliability of current and advanced NDE methods and procedures so that the currently approved code and guide procedures can be either validated or updated and (2) the basic tech-3 niques for ultrasonic test, eddy current, and acoustic emission for continuous l.
monitoring are upgraded through development studies and proved in realistic field studies, in cooperation with commercial ISI vendors and utilities, using operating reactors and components where possible.
The first approach employing round-robins is especially illuminating because it is possible to quantify the reliability of techniques currently called out in the ASME Code or guides and those employed in advanced methods.
The round-robins have included piping of wrought stainless steel, centrifugally cast stainless steel, and clad carbon steel. Thermal fatigue cracks were emplaced in all three types of pipe material, while IGSC cracks were also included in some of the wrought stainless steel pipes.
Piping round-robins are continuing, especially on wrought stainless steel piping with IGSC cracks in typical but hard-to-inspect locations and in cast stainless steels.
This work is conducted in cooperation with OECD. Other round-robins either being conducted by NRC or in which NRC is participating include those on plate and nozzles wherein defects have been implanted in realistic situations or wherein the defects are true manufacturing defects.
This approach has already yielded a series of recommen-dations for changes to improve the code procedures for ultrasonic testing; it has also yielded a data base that permits a valuable revision of the guides currently approved for ultrasonic inspection.
1 Regarding the second approach, a key validation tool for eddy current steam generator tube inspection is the retired-from-service steam generator discussed l
in Section 1.2.
Here, accurate knowledge of the flaw sizes and types and of the extent of degradation measured in tubes after removal from the generator provides the ultimate means to evaluate and validate the inspection method by comparing the actual flaw to the in situ eddy current inspection results.
Continuous monitoring to detect the onset of cracking or leakage through use of acoustic emission is validated by large scale pressure vessel cyclic crack growth tests and by studies on components of actual operating reactors.
A critical part of the program is validation of state-of-the-art ultrasonic test methods fnr both detection and evaluation of flaws in vessels, piping, and nozzles.
Automated systems for real-time inspection and evaluation of flaws in these environments are validated following development in mockups such as at the PISC plates and the EPRI NDE Center and also in the field in operating reactors when appropriate opportunities arise.
Because of the validation and accuracy achieved through these means, licensing criteria and code or guide i
j procedures can be drawn up and used with assurance of improved reliability.
l A most important use of the research results is as the basis for criteria for qualification of personnel, equipment, and procedures, especially for ultra-sonic inservice inspection of piping and other primary system components.
It is because of the insights gained into the effects on detection and evaluation 1-18
. ~.
l 4
reliability of different inspection procedures and use of equipment that such qualification critoria can be set out.
The major research products will be:
1.
a.
Recommendations for. improvement of ASME Code,Section XI, rules for inservice inspection'of nuclear power plant components, to improve the reliability of required inspections (1987-1988).
b.
Completion of evaluations and establishment of reliability of currently practiced and advanced eddy current and other NDE methods, using retired steam generator (1987).
Validation of improved SAFT-UT (synthetic aperture focusing technique c.
for ultrasonic testing) flaw detection and evaluation method in field tests to obtain accurate flaw data for licensing decisions on piping, thick sections, welds, and multimetal joints (1987).
d.
Code case recommendations for acceptance of the improved SAFT-UT and eddy current methods for flaw evaluation and detection and for continuous acoustic emission (AE) monitoring (1987-1988).
(Also applies to Need 2.)
3 e.
Improved inspection plan for implementation in licensing actions for inservice inspection of steam generator tubing (1987).
l f.
Code acceptance of unified set of inspection requirements for piping and vessels based on NDE flaw detection reliability, component mate-rial properties, and service conditions to ensure a suitably low failure probability (1987-1988).
g.
Capability in ultrasonic testing, AE, eddy current, and related NDE methods for prompt reaction to inspection problems (1987-1989).
(Also applies to Need 2.)
2.
Code acceptance of continuous AE monitoring for cracks for licensing use where improved monitoring is necessary or conventional methods cannot be used (1988).
3.
Development and validation of NDE techniques for detection and charac-terization of service produced flaws and material property degradation in reactors under consideration for extension-of-service licenses after initial licensing period has expired (1988-1991).
E 4
l i
1-19 6
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2.
EQUIPMENT QUALIFICATION This program will study the methods used for qualifying equipment used in nuclear power plants. Methods will be validated and new methods developed as appropriate to ensure that qualification test results reported by applicants and licensees provide a basis for licensing decisions that ensure protection of the public health and safety.
The elements discussed in this chapter include environmental qualification of mechanical equipment and dynamic qualification of e?nctrical and mechanical equipment.
Equipment qualification needs will be assessed in light of source term changes resulting frcm ongoing efforts.
2.1 Qualification of Mechanical Equipment (Environmental) l This research will provide the technical basis for developing requirements for environmental qualification of mechanical components.
Environmental parameters include temperature, pressure, humidity, radiation, chemicals, and submergence.
l They do not include consideration of dynamic loads whether these originate from outside the equipment (e.g., seismic or other transmitted vibration) or from inside the equipment (e.g., dynamic effects from process flow).
Dynamic loads are addressed in Section 2.2.
2.1.1 Major Regulatory Needs and Their Justifications 1.
Determination of the environmental parameters affecting the ability of the equipment that is required to perform a safety function during and follow-ing design basis events, to be the basis for licensing decisions and for assessing qualification programs submitted by applicants and licensees (1987).
Justification:
Since mechanical equipment will be subjected to many different environmental parameters, it is necessary to determine which.
environments may affect the safety function of the equipment.
2.
Evaluation of proposed methods of qualifying equipment for design basis events at new and operating plants and those under construction, to be the basis for licensing decisions and the development of regulations or regu-latory guides that endorse national consensus standards (1987).t Justification:
Evaluation of equipment qualification methods is needed so the staff can assess vendor and utility submittals.
Currently, standardized qualification methods may be deficient for some mechanical equipment.
Thus, an independent, unbiased evaluation of various methods must be performed by the NRC.
2.1.2 Research Program Description Those environmental parameters that are significant in affecting the equip-ment's functional capability will be studied to determine if the assumptions currently used concerning the environmental loads are correct and to determine if there are any synergistic effects when those loads are combined, tA modification of the regulations may result from the research.
2-1
The environmental effects are of concern only for limited subcomponents of mechanical equipment such as seals, gaskets, and packing.
The technical bases for evaluating the environmental effects on mechanical equipment will come from the program described in Section 1.4, " Electrical and Mechanical Components."
The major research products will be:
1.
a.
Determination of station blackout effects on main coolant pump seals and the contribution to resolving Generic Issue B-23, " Reactor Coolant Pump Seal Failure" (1987).
b.
Identification of failure modes in pumps and valves for subcomponents such as gaskets, seals, and packings using existing experience data (1987).
2.
Modification or development of standard to include methodology for qualifying main coolant pump shaft seals (1987).
2.2 Dynamic Qualification of Equipment This research will provide the technical basis for developing the qualification requirements involving dynamic loads whether they originate outside the equip-ment (e.g., seismic or other transmitted vibration) or inside the equipment (e.g., dynamic effects from process flow) for electrical and mechanical equip-ment.
It includes environmental loads to the extent that they may be combined with the dynamic loads.
Also included is research on extrapolation, character-ization of loads, load sequencing, load combinations, margins, uncertainties, life extension, and qualification by testing and analysis.
2.2.1 Major Regulatory Needs and Their Justifications 1.
Determination and characterization of those loads affecting the ability of the equipment that is required to perform a safety function during and following design basis eveats, to be the basis for licensing decisions and for assessing qualification programs submitted by applicants and licensees (1987-1988).
Justification:
Since mechanical and electrical equipment will be subjected to many different loads during the life of a plant, it is necessary to determine whR characteristics of the loads may affect the safety function of the equipment and the uncertainty of the magnitude or level of the loads to be simulated during qualification.
The research will also identify areas in which the uncertainties in defining the dynamic parameters may be beneficially reduced.
2.
Establishment of data on equipment responses, failure modes, and margins, to be the basis for ensuring the seismic capability of mechanical and electrical equipment in new and operating plants (1987).
Justification:
Establishment of this data base is needed for assisting the staff in judging structural integrity and functional operability of equipment claimed to have been seismically qualified by applicants and licensees.
J 2-2
3.
Determination of safety margins that are available in the existing mech-anical and electrical equipment design against the safe shutdown earthquake (SSE) for new and operating plants (1987).
Justification:
Determination of the safety margins is needed'for assess-ing equipment capability under the effects of earthquakes, including those with magnitudes greater than the design basis earthquake.
4.
Evaluation of proposed methods of qualifying equipment for design basis events in new and operating plants and those under construction, to be the basis for licensing decisions and the development of regulations or regulatory guides that endorse national standards (1987).t Justification:
Evaluation of methods of mechanical and electrical equip-ment qualification for dynamic loads are needed for the staff to assess vendor and utility submittals.
Currently, standardized qualification methods may be deficient for some mechanical equipment.
Thus, independent, unbiased evaluation of various methods must be performed by the NRC for these components.
5.
Establish criteria for determining what qualification methods are accept-able for new plants and those under construction, to be the basis for a regulatory guide (1987).
Justification:
Prudent acceptance criteria for mechanical and electrical equipment qualification methods for dynamic loads must be developed.
These criteria must account for uncertainties in definition of the dynamic loads and the qualification methods, yet they must provide a measure of reduction in risk over mechanical and electrical equipment qualified to other less appropriate criteria.
6.
Develop methods and requirements necessary for determining whether the life of specific equipment and components can be extended beyond 40 years, for use in making licensing decisions as to whether operation beyond 40 years should be approved (1990).
Justification:
Current thinking is that some operating plants and equipment can be safely operated beyond 40 years.
The reasons for this include overconservatisms in the original designs, the low loading histories experienced by the plants and equipment during the 40 years,.
and the assurance that evaluation techniques can adequately predict the existing conditions of the safety equipment.
Guidance for identifying those components that need not be replaced will be necessary to ensure that the safety function of those components will not be jeopardized during future design basis events.
The development of this guidance as well as requalification requirements will provide the basis for modifying standards currently used by the industry.
2.2.2 Research Program Description This research program will provide the technical basis to evaluate the qualifi-cation procedures for mechanical and electrical equipment subjected to dynamic loads. This program consists of experimental and analytical efforts that will provide data and information for determining and establishing margins for com-ponents such as valves, pumps, motors, batteries, and electrical controls when tA modification of the regulations may result from the research.
2-3
subjected to postulated accidents.
The loading characteristics, modes, and levels of failure (if any) for these components will be identified and will aid in providing criteria for evaluating equipment in plants currently being licensed and new plants.
Included with the above effort is that related to determining the characteristics and the effects of long-term system loads on the operability of pumps and valves.
Acceptable methods for accounting for these loads in the qualification process will be developed.
In addition, methods and criteria for requalifying equipment for use in plants licensed to cperate beyond the currently allowed 40 year life will be developed.
The important considerations regarding the dynamic qualification of equipment such as valves and electrical controls by test and by analysis will be identi-fied.
The question as to when a component can be qualified by analysis rather than by test will be answered.
In addition, modeling techniques and methods for simulating the effects of nonlinearities in analytical models will be devel-oped. Methods for identifying and determining the flow-induced force charac-teristics will also be developed.
Areas of importance for-dynamically qualifying categories of equipment by test rather than by analysis include accounting for the influence of system inter-actions and the simulation of component foundations, including methods for ensuring input due to cross-coupling.
Experimental studies sponsored by other NRC research programs as well as foreign and domestic agencies will be used to the maximum extent possible.
Information exchanges with EPRI in the area of equipment qualification are being explored.
High-level vibratory tests at the HDR facility in the Federal Republic of Germany include evaluation of active equipment under simulated seismic condi-tions.
The Japanese Ministry of International Trade and Industry (MITI) has expressed interest in the test results thus far obtained, and information exchanges are being planned.
Preliminary information exchanges with the Atomic Energy Control Board of Canada have begun.
Interactions with that agency are expected to continue as results of their environmental qualification program become available.
During the experimental parts of the program, where possible, information pertaining to containment leak integrity will be provided to the severe accident research program.
Interaction with various committees that develop component qualification standards has begun and will continue when appropriate milestones in this pro-gram are achieved.
The major research products will be:
1.
a.
Criteria and guidelines for qualification by test and analysis for valves (e.g., containment boundary valves), motors, and electrical controls (1987).
2-4
b.
Identification of current qualification criteria for specific valves in plant safety-related piping systems in which failure to close against high flows resulting from downstream pipe breaks would degrade plant safety (1987).
c.
Determination of whether isolation valves in high-energy safety-related piping systems will close against high flows from downstream pipe breaks (1987).
2.
Establishment of a data base of equipment responses and failure modes to determine whether failures of equipment such as valves, pumps, motors, and electrical controls may be due to inadequate qualification requirements (1987).
3.
Evaluation of safety margins available in mechanical and electrical equip-ment such as valves, pumps, motors, and electrical controls subjected to the SSE (1987).
4.
Development and evaluation of methods that are acceptable for dynamic qualification of mechanical and electrical equipment such as valves, pumps, motors, and electrical controls (1987).
5.
Development and evaluation of criteria for determining acceptable methods for quantifying loads on mechanical and electrical equipment such as valves, pumps, motors, and electrical controls (1987).
6.
Development of methods and criteria for requalifying mechanical and electrical equipment to ensure its safety function in plants licensed to operate beyond 40 years (1990).
l 2-5
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l 3.
SEISMIC RESEARCH Earthquakes are among the most severe of the natural hazards faced by nuclear power plants. The bases for NRC's seismic research needs are three issues dealing with seismic hazard, seismic risk, and seismic margin.
Since the U.S. Geological Survey (USGS) clarified its position on the Charleston earthquake, the seismic hazard of nuclear power plants on the eastern seaboard of the United States has become a major regulatory issue. The question is whether an earthquake of the magnitude of the Charleston earthquake could occur elsewhere on the eastern seaboard, placing a nuclear power plant at hazard.
Another issue is the public risk associated with seismic events in excess of the current design basis, the safe shutdown earthquake (SSE).
Should this risk turn out to be large, the technical bases for potential changes in the regula-tions may be needed.
The third seismic issue is the margin inherent in the seismic design of older plants.
Frequently, the NRC is faced with decisions relating to the seismic design of operating plants.
New information available on seismicity, soil-structure interaction (SSI), and plant seismic response would result in changes in the predicted loads on structures, systems, and equipment.
These changes must be compared to the inherent margin of the original design before a decision can be made.
3.1 Seismic Hazard 3.1.1 Major Regulatory Needs and Their Justifications 1.
Data concerning seismic source zones in the Eastern United States (east of the Continental Divide), including Charleston, New Madrid, New England, and others, and data concerning seismic energy propagation in the East needed for the analysis of seismotectonic provinces as required by Appendix A to 10 CFR Part 100, to be the basis for amendments to Appendix A to 10 CFR Part 100 and for develaping regulatory guidance (1987).t Justification:
Except for the New Madrid seismicity, the distribution of seismicity in the East, is not well defined.
No working hypothesis for the cause of the seismicity is generally accepted by the geoscience commu-nity.
In addition, continued seismic monitoring in the Central and Eastern United States is needed by the NRC to support its regulatory functions with regard to operating nuclear power plants.
Therefore, the NRC will continue to support its regional seismographic networks through 1992, during which time the USGS will, in cooperation with the NRC, install and initiate operation of a new seismographic network in the Central and Eastern United States.
tA modification of the regulations may result from the research.
3-1
2.
An information base for the development of site-specific response spectra, to be the basis for amendments to Appendix A to 10 CFR Part 100 and for developing regulatory guidance (1990).t Justification:
The recent earthquakes in New Brunswick, New Hampshire, and Arkansas have generated important strong-motion records that for the first time provide a significant opportunity to compare real data with theoretical ground motion and attenuation models for the Eastern United States.
Analysis of these records will address important regulatory questions concerning the interpretation of this, type of record and its use in licensing decisions.
3.
Methods for handling the uncertainties in assessing the potential risk from seismic hazards, including such topics as the applicability of using the historic method of seismic hazard analysis, the impact of the 1982 New Brunswick earthquake sequence, and the verification of Holocene movement on the Meers fault, to be used to revise current siting regulations (1990).t Justification:
The current seismic siting regulations do not provide guidance on how to handle the uncertainty associated with the licensing decisions and judgments being made at the forefront of a rapidly developing science, seismology.
Decisions must be made based on the best available data that are steadily being updated.
This program is designed to develop statistical or probabilistic tools to aid the decisionmaking process.
4.
Improved data base and analjsis techniques for predicting soil failure, including soil liquefaction, particularly for small or moderate changes in the acceleration value at which the design spectrum is anchored, to be used for regulatory guidance (1987).
Justification:
The recent hig5 acceleration records from New Brunswick, New Hampshire, and Arkansas and the potential for large " anchor point" accelerations because of the Charleston earthquake issue may reduce the safety margins associated with cucrent soil-failure prediction techniques.
5.
Verification of methods for predicting seismic soil settlements (1987-1988).
Justification:
The consequences of soil settlement, liquefaction, or the soil failures at high earthquake levels (above SSE) could be a significant contributor to overall risk.
Current probabilistic risk assessments (PRAs) do not adequately address this problem, and at the present time there are no verified methods for estimating seismic soil-settlement.
3.1.2 Research Program Description Uncertainty in seismic hazard analysis is a fundamental issue.
The strategy to reduce this uncertainty involves three programs:
(1) development of a better seismic zonation through studies of the causes of earthquakes in the Eastern United States, (2) determination of more accurate seismic wave attenua-tion relationships, and (3) development of better data and models of site-specific spectral response.
tA modification of the regulations may result from the research.
3-2
I The program to establish a better seismic zonation in the Eastern United States 1
is directed at determining the cause of the seismicity in the East.
This program consists of monitoring the seismicity in the East through a series of seismographic networks, studying the crustal structure in critical areas (such as Charleston, S.C., Moodus, Conn., and the Ramapo fault area, N.Y. and N.J.),
crustal stress measurements, studies of recent crustal movement, and detailed evaluation of features such as the Meers fault in Oklahoma.
To carry out this program in the earth sciences effectively and efficiently, the NRC is actively interacting and cooperating with other Federal, State, and private agencies l
such as the USGS, State geological surveys, and the Electric Power Research l
Institute (EPRI).
Other projects involve the more precise placement of seismic i
instrumentation and the analysis of data to establish seismic wave attenuation i
relationships and to limit uncertainties in ground motion.
The NRC will continue to support the operation of some of its seismographic networks through FY 1992 and is negotiating with the USGS to establish a new national seismo-graphic network in the Central and Eastern United States that will be fully supported and operated by the USGS.
A network of strong-motion seismographs has been established to gain informa-tion on site-specific response and attenuation relationships.
The strong-motion data being collected on Taiwan at the EPRI test facility will be used in 4
a cooperative assessment of analytical methods to predict the effects of earth-i quake motion.
A representative piping system has been constructed on an instru-mented test platform in the northern part of Taiwan, an area frequently subjected to strong seismic ground motion.
Data analyses and theoretical studies of strong motion are being conducted for the Eastern United States in cooperation with the USGS engineering seismology group.
The Army Corps of Engineers has studied soil-failure models and concluded that the DESRA code is the best candidate for validation.
Validation experiments are being conducted at Cambridge University.
The major research products will be:
1.
Seismographic network data, geologic data, including maps, boring logs, paleoliquefaction feature details, and geomorphic and geodetic data for determining neotectonic aspects of the Eastern United States, used on a day-to-day basis by licensing staff, by staff involved in PRAs, by the seismic hazard characterization project, and for rulemaking decisions and engineering research projects and geophysical data for determining crustal structure in areas of suspicious geologic structures (1987-1991).
2.
Techniques for calculating site-specific response spectra (1989).
i 3.
a.
Probabilistic sensitivity study of probable ground-motion dependence on the various proposed causes of seismicity in the Eastern United States (1987).
b.
Results of field investigations of the Meers fault in Oklahoma (1989).
3-3 E - - -
4.
Validation of the DESRA code and recommendations regarding the use and criteria for the use and importance of SSI analytical methods for layered soil, high water tables, and liftoff effects (1987).
5.
Determination of seismic risk contribution from dam and embankment.
failure (1988).
3.1.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
As a result of budgetary reductions for FY 1987, a program to determine the contributions of dam and embankment failures to seismic risk has been deferred.
Probabilistic techniques to evaluate the effects of earthquakes equal to or larger than the SSE are needed for reviews of the geotechnical features of Seismic Category 1 dam and embankment structures to ensure that they can perform their functions during the life of the nuclear plant.
3.2 Seismic Risk 3.2.1 Major Regulatory Needs and Their Justifications 1.
Validation of current seismic PRA' methods (1987-1989).
Justification:. The Seismic Safety Margins Research Program (SSMRP) and industry methodologies for estimating seismic risk are fairly new and have not been subjected to experimental validation.
The seismic hazard and fragility data rely heavily on expert opinion, and critical system model-ing assumptions are made in the risk analyses.
Validation will increase our confidence in seismic PRA methods and their effectiveness so that they may be used in the regulatory decisionmaking process.
2.
Development of a methodology for flood probability estimates based on stochastic relationships between deterministic data and flood probabilities (1988-1990).
Justification:
Deterministic methods of evaluating severe hydrological and hydrometeorological events (i.e., floods, droughts) do not support risk or safety margin assessments.
Methods for making reasonable and defensible flood probability estimates and developing flood hazard curves are needed by tlw NRC staff to replace interim procedures using default values.
3.
Investigation of the levels of uncertainty in current tornado wind speed estimates based on calibrations of structural damage to inferred wind speeds (1988).
Justification:
A determination of the error bands and quantification of uncertainty estimates in the tornado data base used by the NRR staff is needed to reduce uncertainties in risk evaluations, including those for operating plants, concerning hazards due to high winds.
4.
Evaluation of real-time atmospheric dispersion and plume rise models for dose projections during emergencies, to determine model accuracy and re-liability, uncertainties, and data requirements under various terrain and meteorological conditions (1988-1990).
3-4
o Justification:
Information concerning the accuracy and reliability of real-time dispersion models used for dose projections within the emergency planning zone, including uncertainties and data requirements, under a variety of meteorological and terrain conditions is needed by the IE and
'NRR staffs for reviews of licensee and applicant emergency response organiza-tions and accident assessment capabilities.
5.
Confirmation of washout coefficients for released, water-soluble partic-ulate and gaseous fission products that determine the fraction of the effluent plume that is scavenged and brought down to the ground as a j
result of natural rainfall occurring along the plume pathway during the course of an accident (1988-1990).
i Justification:
A confirmation of the values for the washout coefficients for released, water-soluble gaseous and particulate fission products due i
to rainfall, which are currently controversial, is needed by the NRC staff to reduce uncertainty ranges in probabilistic risk assessments and for emer-gency response applications.
4 3.2.2 Research Program Description In the past, a general lack of instrumentation has prevenf.ed the correlation of earthquake damage to specific input spectra.
At present, a program instituted by the U.S. Bureau of Reclamation associated with the California Department of Water Resources has caused a significant increase in the place-ment of seismic instrumentation.
This should prove useful for this validation program.
An example of instrumentation that was strategically located is the Pleasant Valley station that provided useful information during a May 2,1983 l
earthquake at Coalinga, Calif.
The techniques and data used for predicting the seismic hazard at sites will be validated through the seismotectonic program (Section 3.'1.2).
Products from this program will also help reduce the large uncertainties now associated with predicting the occurrence of earthquake levels of interest in seismic PRAs.
The SSMRP has been the most comprehensive effort to date in the seismic risk field.
The products of this program will serve to benchmark other PRAs that consider seismically induced accidents. While a number of comparisons with other analytical methods have been performed under the program, an increased i
effort to independently validate the SSMRP methodology using experimental'and observed earthquake data is being pursued.
j
.In 1984, an expert panel on seismic risk validation formulated recommendations for research needed to verify or improve methods of structural response and failure prediction that are in use in current seismic risk studies.
The expert panel on quantification of seismic margins has also identified seismic PRA validation needs.
In 1985, a program plan for seismic risk validation was completed. This plan will coordinate ongoing research and will develop new structural research to meet the needs recognized by these expert panel, Two important elements of this plan were started in 1984 with the establishment of j
NRC cooperation in the high-level HDR shake tests in West Germany and in EPRI's containment model experiments in Taiwan.
Also, an agreement is currently being l
negotiated between Japan's Ministry of International Trade and Industry (MITI) and the NRC to conduct a validation test on a PWR primary coolant loop.
i 3-5 l
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l The HOR facility is a modification of a superheated steam reactor that was decommissioned and modified for research in 1973.
Phase I tests involving experiments on materials engineering, thermal hydraulics, and mechanical and earthquake engineering were conducted during 1975-1983.
The primary focus of
)
these experiments was comparison of predictions by analytical models with experiments.
The Phase II experiments, during 1984-1988, are similarly motivated and will examine higher levels of response where damage to structures, systems, and components is expected.
NRC participation in the seismic tests will involve providing predictions for the response of structures and piping systems excited by shakers.
One series of experiments, in which the containment building will be excited by a large shaker, is planned for 1986.
Another series in which piping systems will be excited into the inelastic range is planned for 1987.
EPRI is constructing, in a seismically active area in Taiwan, a model about 1/4 the size of a concrete containment.
EPRI will also install instruments in the model and in vertical and horizontal arrays in the vicinity of the model and will record earthquake responses over a 5 year period.
NRC will perform 4
low-level vibratory tests of the model to provide baseline data on modal parameters.
Future NRC effect will consist of providing analytical models to predict SSI effects for the recorded earthquakes and of comparing predictions with observations.
A massive testing effort under the sponsorship of the Japanese MITI was started in 1982 and will continue through 1988.
NRC's main interest in this test series is to determine the ability of analytical methods to predict the onset of component damage under very large earthquake motions.
To that end, the NRC is negotiating for tests to be performed after " proving tests" have been completed on one specimen, a PWR piping loop model.
The NRC and the Japanese anticipate that the specimen will respond elastically to the proving tests and will be in an undamaged state. The tests in which the NRC will cooperate would involve increasing the excitation (within the limits of the table) and modifying the specimen to induce inelastic response.
The PVR piping loop test will be performed in 1987.
Two aspects of structural behavior are important in evaluating the safety of nuclear power plants.
The first aspect deals with predicting the internal forces resulting from applied external loads and constraints.
The second deals with evaluating the strength or capacity of the structure.
Experimental 1
studies have been conducted in each of these areas.
The specific objective of the research is to evaluate sources of experimental data that relate to both aspects of the problem.
The data collected on these experiments will be reviewed, and a program plan will be developed and implemented to produce structural benchmark problems.
The benchmarks developed will be used by the i
licensing staff to validate licensee methods for predicting the behavior of i
safety-related structures under seismic loads.
A feasibility study on the adaptation of a Canadian Atomic Energy Control Board study and other existing methodologies of comparing presently used deterministic techniques to stochastic techniques for estimation of design basis floods for nuclear power plant siting will be performed.
i 1
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Also presently used NRC staff methods relating probable maximum events (i.e.,
probable maximum flood (PMF), probable maximum hurricane (PMH), etc.) to j
probabilistic-based events and estimating the degree of error in the developed relationship will be evaluated.
Based on the results of the feasibility study, a second study will be performed to review and assess the problem of determining severe event probabilities up to the probable maximum event (i.e., PMF, PMH, etc.) and the various methodologies proposed in the feasibility study and in the technical literature.
The merits and deficiencies of the proposed methodologies will be evaluated.
Also, the current NRC staff methods for determining the probabilistic relationship to deterministic-based severe flood events against the proposed methods will be assessed and alternative methods or modification of existing methods will be recommended.
The levels of uncertainty in current tornado wind speed estimates based on calibrations of structural damage to inferred wind speeds will be investigated.
The feasibility of quantifying uncertainties in tornado hazard curves through snalysis of known sources of error will be determined.
Dispersion models for emergency response applications will be evaluated by com-paring their projections of plume location, shape, arrival time, and concentra-tions with data from field tests to determine the validity and accuracy of each model under various terrain and meteorological situations.
The most accurate and reliable models will be identified along with the meteorological data input that best ensures the accuracy of the model projections.
Existing methodologies for characterizing plume rise will be evaluated to determine which are most applicable to emergency situations at nuclear facilities, including fires.
Investigations to confirm washout coefficients for fission product plumes under natural rainfall occurrences during the course of an accident will build on information obtained from previous wind tunnel experiments and will use data obtained from small-scale field tests.
The major research products will be:
1.
a.
Nonlinear response data from HOR shake tests (1987).
b.
Earthquake response data from EPRI's scale-model containment in Taiwan (1987-1990).
2.
Report describing the technical base to be used to develop guidance for an acceptable methodology for making flood probability estimates and for developing flood hazard curves for use in risk assessments and safety margin evaluations (1989).
3.
Report defining uncertainties and resulting error bands associated with tornado wind speed estimates inferred from structural damage surveys and quantifying uncertainties in tornado hazard curves (1989).
3-7 n
l 4.
Based on comparisons of model predictions with results from field tests, information concerning the suitability and reliability of various atmos-pheric dispersion models for real-time emergency response applications
-related to plume characterization and dose projection within the emergency planning zone (1990).
5.
' Final report providing information concerning appropriate values for 1
fission product plume washout coefficients under rainfall or other natural precipitation events (1990).
3.2.3 Deferred /Unfunded Research Needs (as of February 1. 1986)
As a result of reductions in the FY 1987 budget, the following programs have been deferred: development of a methodology for flood probability estimates, investigation of the levels of uncertainty in current tornado wind speed esti-mates, validation of real-time dispersion and plume rise models for emergency-dose projections during accidents, and confirmation of washout coefficients for fission product plumes that are impacted by rainfall.
3.3 Seismic Margins 3.3.1 Major Regulatory Needs and Their Justifications 1.
Methodology and procedures to systematically assess the ability of nuclear power plants to withstand earthquakes greater than those of the i
design bases (1987).
Justification:
Nuclear plants traditionally have been designed to be safely shut down after a design basis earthquake using conservative methods and criteria. Although ensuring design adequacy, this deterministic approach does not give a true picture of the actual margin to failure, nor does it provide enough information to make realistic estimates of seismic risk.
From a licensing perspective, there is a continuing need for consideration of the inherent quantitative margin against earthquakes, among other reasons because of the changing perceptions of the seismic hazard.
Any reevaluation of plant seismic resistance using methods similar to current design practice would require a rigorous reanalysis of systems, structures, and components even if there were only'a slight increase in the earthquake level.
A desired alternative is the use of a seismic margin methodology that relies both t
on seismic PRA information to focus the number of systems being reevaluated and on test and earthquake experience data to minimize the amount of analysis needed to assess the seismic resistance of structures and components.
2.
Data to predict the nonlinear response and failure modes of nuclear shear wall structures (1987-1990).
Justification:
Recent seismic PRAs indicate that structural failures leading to failure of the equipment housed or supported are dominant risk contributors.
Our lack of appropriate test or earthquake experience data has resulted in large uncertainties associated with structural fragilities.
It is also necessary to more clearly understand the nonlinear response of shear wall structures so that the high seismic load input to equipment can be better defined.
3-8 J
m
4 3.
An improved seismic fragility data base for mechanical and electrical equipment (1987).
Justification:
The fragility data used in current PRAs and in future seismic margin evaluations rely heavily on expert opinion and thus are subject to large modeling uncertainties.
A better understanding of equip-ment failure modes and levels and the parameters that control them is needed as a basis for equipment qualification decisionmaking and for standards leading to more balanced plant design.
4.
A basis for improving piping and piping support design criteria (1987-1988).
Justification:
Current design criteria for piping systems lead to an imbalance in design between dynamic and other loadings.
Emphasis on design is now directed toward analytical rigor that adds cost without completely ensuring good design practice.
3.3.2 Research Program Description The Seismic Category I structures program will provide analytical and experi-mental data assessing how the p cameters used in the design of safety related equipment and noncontainment structures are affected by earthquake loads above the initial design level. These parameters include, but are not limited to, floor response spectra, structural frequencies, accelerations, and displace-ments.
Sensitivities of these design parameters to changes in internal and 1
external wall configurations, design practices, and magnitude and duration of j
seismic input motion will be determined as the model configurations are subjected to quasi-static and seismic tests of increasing magnitude to cause elastic and inelastic responses.
These sensitivity tests are necessary because the auxiliary buildings, diesel generator buildings, etc., as placed in the various nuclear power plants, are very plant specific.
In order to gain an insight into the effects of additional walls, different numbers of floors, or different sizes of buildings, several parameters are varied so the results can be applied to the various configurations that are actually in operating plants.
This research will support seismic margin objectives by providing analytical methods and experimental data to support improvements and reduce uncertainties in the current seismic design criteria.
This research will also verify the adequacy of assumptions and subjective information used in probabilistic structural risk analyses by providing failure modes and levels for the majority of configurations tested.
In some limited cases, the model may be so large that the experimental facility would not have the capability to fail it.
i The program for reliability analysis of nonlinear behavior of concrete structures will enable the NRC staff to generate fragility curves for PRA i
studies and to verify results of fragility analyses in PRAs.
This program will also provide information for seismic margin studies and could be used to assess the effects of a change of design basis such as increased seismic loads.
In 1985-1986, the NRC piping research on dynamic load criteria was primarily directed toward addressing the separate research recommendations made by the NRC Piping Review Committee.
In 1987 through 1988, the research will be 3-9 i
focused more toward two general areas.
The first deals with the development and implementation of potential ASME Code changes that would provide better and more realistic dynamic load design criteria. The EPRI/NRC Piping and Pipe Fitting Reliability Program will play a primary role in this effort.
The second general area will deal with a general simplification of piping and piping support design rules using the earthquake and service experience of nonnuclear facilities as guidance as well as new data on failure modes and response behavior of piping.
The objective of this latter effort is to lessen the unnecessary burden of piping and pipe support analysis on new designs and reevaluations by developing rules that emphasize good design practice and deemphasize numerical calculations.
Information on seismic fragilities of mechanical and electrical components will be obtained by testing and by assembling, analyzing, and interpreting existing component fragility data.
This information will be used to improve seismic PRAs and to obtain better estimates of seismic margins by using more realistic test-based component fragilities.
The impact of scaling, aging, anchorages, and other parameters such as vintage, manufacturer, and construction practices will be evaluated.
The Seismic Design Margins Program (SDMP) was developed to address concerns about the effects of earthquake levels greater than those considered in the design basis of nuclear power plants.
The SDMP has used a newly expanded information base concerning the behavior of plant systems and components during i
earthquakes greater than design levels.
In addition to the NRC SSMRP, several industry-sponsored seismic PRAs have become available and have been reviewed.
The earthquake experience data of equipment and piping in nonnuclear facilities have been comprehensively studied and were recently used to evaluate the seismic resistance of similar nuclear components.
Numerous high-level dynamic tests of structures and equipment have also provided new information.
An NRC-sponsored panel of expert consultants has completed a review and synthesis of this infor-mation base and used it in forming a recommended approach to assess the seismic margins of nuclear power plants.
l The seismic margins review procedures developed by the expert panel make use of a screening approach and the concept of determining the level of "high confid-l ence of low probability of failure" for important components and systems.
Seismic PRA results have been used to focus the review of the systems to only those considered most critical to plant shutdown following a large earthquake.
i A success-oriented approach has been used in evaluating the seismic resistance of structures and equipment.
This best uses experience data for the levels that are believed most important for the review.
l The seismic margin review procedures will be demonstrated in 1986 by means of a trial plant review of a PWR.
In 1987, these procedures will be updated to reflect the insights gained from this trial review and to account more thoroughly for the systems behavior insights gleaned from BWR seismic PRAs, which only recently became available in quantity.
A seismic margin review of a BWR will be started in 1987.
3-10 A
The major research products will be:
1.
a.
Refinement of seismic margin review guidelines and recomr.endations based on initial plant review and findings from BWR seismic PRAs (1987).
b.
Trial seismic margin review of BWR (1987).
2.
a.
Determination of sensitivity of structural behavior (i.e.,
fundamental frequency; damping; and response spectra, accelerations, and displacements at various floor levels) to changes in internal and external wall configurations and increased earthquake magnitude beyond the initial design basis resulting in elastic and inelastic j
behavior (1987-1989).
i b.
Failure modes and failure level data for various concrete configura-tions (1987-1989).
c.
Reliability analysis method and stochastic nonlinear analysis method for concrete containments and shear walls subjected to multiple str.iic and dynamic loads that produce nonlinear behavior (1988).
d.
Revisions, as necessary to Regulatory Guides 1.61, " Damping Values for Seismic Design of Nuclear Power Plants," and 1.122, " Development of Floor Design Response Spectra for Seismic Design of Ficer-Supported Equipment or Components" (1988).
l 3.
a.
fragility data from component testing (1987).
b.
Risk comparisons among piping systems designed to various proposed new criteria (1987).
c.
Recommendations for damping values to be used in piping analysis, including the effect of piping system equipment and its effect on system analysis and damping (including composite damping) (1987).
4.
a.
Bases for improved ASME Code piping design rules for dynamic loads (1987).
b.
Bases for simplified piping and piping support design rules (1988).
3-11
4.
REACTOR OPERATIONS AND RISK This chapter describes the research being carried out to support the develop-ment and application of probabilistic risk assessment (PRA) methods to identify those elements of reactor operations that are the most significant contributors to risk and the causal factors associated with them and to permit comparative evaluations to be made of the risk-reduction effectiveness associated with current or proposed regulatory requirements.
This work includes the develop-ment and trial use of models, methods, procedures, and other analyses required to support Commission decisions on a broad range of critical issues relating to power reactor safety and the acquisition of data required to support these applications.
This program is divided into three elements corresponding to the principal topical areas within this program, i.e., risk methodology and data development, operational safety reliability, and severe accident risk assessment.
The topic of severe accidents is also discussed in Chapter 6.
4.1 Risk Methodology and Data Development Research in this element is directed toward developing, testing, documenting, and, to the extent possible, validating methods for estimating the probabil-ities and consequences (risk) of reactor accidents and toward identifying the major contributors to, and the uncertainties associated with, such estimates.
Data obtained from operating nuclear power plants and from experimental research programs are analyzed, evaluated, and incorporated into these models and related data bases.
The information developed in this element provides one important factor in evaluating the need to take regulatory action to correct identified safety issues in operating plants and provides methods for analyzing the costs and benefits of proposed resolutions of such issues.
4.1.1 Major Regulatory Needs and Their Justification 1.
Documented methods for the systematic identification and evaluation of principal reactor accident sequences and their precursors, to support a range of regulatory decisions regarding severe accidents (e.g., accident risks under NEPA) and to provide a consistent basis for evaluating reli-ability assurance programs and operator actions to mitigate accidents (accident management) (1987).
2.
Techniques for incorporating the contribution of common-cause failures, including fire and systems interactions, into PRA methods to support the proposed plan for implementing the Commission's severe accident policy and other PRA related regulatory issues and to complement and extend PRA procedure guidance provided in NUREG/CR-2815 (1987-1991).
3.
Methods for quantifying the effects of severe natural phenomena, e.g.,
seismic activity (1987) and external floods (1987-1988), to include develop-ment of screening procedures and consideration of effects of secondary failures, impacts of mitigating systems, effectiveness of recovery actions, and evaluation of potential multiple accident initiators triggered by the same external event.
4-1
4.
Methods for acquiring and applying reliable human error data for use in making quantitative and qualitative assessments of nuclear power plant operator and maintenance personnel reliability as part of the PRA process (1987-1990).
5.
Procedures and models for adequately assessing the impact of human error on overall plant risk and for providing insights into the causes of enan-man and man-machine errors and identification of methods for their elimin-ation (1987-1990).
6.
Methods to assess and interpret the quantitative uncertainties in PRA modeling and data in support of regulatory decisionmaking, including importance measures and sensitivity analyses (1987-1991).
Justification of Above Needs: Quantitative assessments of the probabilities and consequences of severe reactor accidents are becoming an increasingly important element in the regulatory decisionmaking process.
The Commission's Policy Statement on Safety Goals includes quantitative risk-based design objectives.
Similarly, policies presently being considered by the Commission (SECY-84-370, "NRC Policy on Future Reactor Designs:
Decisions on Severe Accident Issues in Nuclear Power Plant Regulations") will require that future Commission decisions regarding the imposition of additional safety requirements on operating plants be justified, among other things, on the basis of the best available evidence (including probabilistic analyses) of the current level of safety of reactors and the potential for risk reduction.
In other areas, PRA techniques and models being developed in this program will become increasingly important tools for assessing the safety significance (and priorities) of many important issues facing the Commission (e.g., the unresolved safety issues),
for evaluating new potential safety issues as they arise, and for evaluating and reviewing industry-sponsored PRAs and PRA applications (e.g., 10COR).
Although methods to assess the likelihood and consequences of reactor accidents are reasonably well developed, a number of sources of uncertainty remain that limit their usefulness.
For example, a substantial amount of research should be performed so that common-cause failures and human errors can be more compre-hensively included in risk analyses.
Research is also planned to strengthen our capability to assess (i.e., to quantify with known error bands) the risk associated with extreme natural phenomena such as external floods and seismic activity.
Modifications are also required in the models used to estimate the consequences of the severe reactor accidents considered in PRAs.
Present accident risk codes that predict fission product behavior have not yet been updated to reflect the results of recent (and planned) experimental research.
Improved quantifica-tion of the uncertainties in likelihood and consequence prediction is also needed to improve the use of PRA in decisionmaking.
In addition to research to imprcve risk methods, there is a corresponding need to collect and analyze data on the performance and failure rates of plant systems and components.
This activity will help to ensure that the results of reactor risk assessments are as accurate and reliable as possible; the analysis 4-2
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I of such data can lead to the identification of incipient safety problems in operating plants, thereby avoiding events that might pose a risk to public i
health such as those that recently occurred at the Davis-Besse and Salem i
stations.
4.1.2 Research Program Description i
This research is being conducted with the ultimate objective of providing i
improved PRA techniques that will facilitate the integration of PRA into the regulatory process.
The program will provide detailed, documented methods that can be used to ensure consistency and completeness in PRA execution.
l These methods will be exercised and tested in an integrated and realistic manner in the Risk Methodology Integration and Evaluation Program (RMIEP),
i 1-which will apply the newly developed methods in a test probabilistic risk analysis.
I Together with the method development activities, research is being conducted to evaluate and interpret data from operating plants.
This program will pro-vide the most up-to-date information on equipment and component failure rates, I
on the root causes of such failures for use in resolving current and future j
safety issues, and on establishing criteria for reliability assurance programs.
I i
The major research products will be:
i i
6 i
1.
Documented review of accident sequences and their likelihood, including accident precursors, using improved methodology and development of simpli-i fled models combined with data bases that will permit evaluation of a wide variety of issues and review of operating history from a risk perspective (1987).
1 2.
A procedure to incorporate analyses of dependent failures, including common j
cause failures such as common maintenance errors, into a PRA (1987-1988).
)
3.
Procedures for incorporating results of simplified Seismic Safety Margin Research Program methodology into PRA methodology (1987-1988).
(To be developed and applied for trial use in the RMIEP).
4.
Computer-based models for developing human error probabilities for selected nuclear power plant operations and support personnel functions (1989).
5.
Procedures for fully integrating human reliability analyses into the PRA process to achieve more adequate assessments of the effects of human per-l formance on overall plant risk (1987-1991); methods for systematically using human reliability analyses and PRA results to identify man-man and 1
man machine interfaces of safety importance at nuclear power plants I
(1987-1988).
r l
6.
Documented validation of integrated risk assessment methods, using improved data on external event frequencies, common cause failure rates, and the j
root causes of system failures; validation to provide quantitative display 1
of uncertainties in such risk assessments (1987-1988).
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l 4.2 Operational Safety Reliability The research programs in this element are directed toward the specific applica-tion of risk-based information to issues relating to the maintenance and improvement of the safety of operating reactors.
It is intended that these programs develop practical tools for making risk-based information available to the licensing and inspection staffs within the context of their day-to-day decisionmaking functions.
In addition, these programs will provide a strong basis for establishing criteria for licensees' reliability assurance programs, including the requirements of an independent NRC audit capability. An informa-tion base will also be developed to assist the staff in assessing the nature and potential consequences of abnormal events in operating reactors and in identifying the effectiveness of potential corrective actions that might be taken by plant operators.
4.2.1 Major Regulatory Needs and Their Justification 1.
Demonstration of the feasibility, usefulness, and cost effectiveness of selected operational safety reliability methods validated by use in actual plant operating situations (1987).
2.
Methods that will provide risk-based information to the licensing staff to assist in evaluating the scope, content, and impacts of changes in operating plant technical specifications and in establishing limiting conditions of operations and design changes (1987-1991); methods that will provide risk-based information to the Office of Inspection and Enforcement (IE) to assist in the development and implementation of modules for operating license inspections that take into account current plant status, analyses of accident sequences, plant operating history, and accident likelihood (1987-1991).
3.
Methods and techniques for applying PRA insights and risk estimates to plants that have not prepared a PRA (1987-1991).
Justification of Above Needs:
Since the Reactor Safety Study (WASH-1400) was published, there has been a strong and continuing interest in applying the insights and information contained in PRAs to specific regulatory needs.
Generic PRA information has been used to support closure of several of the unresolved safety issues such as anticipated transient without scram (ATWS), station blackout, and pressurized thermal shock.
However, addi-tional work is required to permit the use of risk-based information in a practical way in a number of other important areas such as licensing and inspection.
Methods must be developed to distill the information and insights contained in the voluminous PRA documents (that basically were written for PRA practitioners) into a form that can be used to support the day-to-day decisions of the staff.
Such risk-based information would be of great assistance in evaluating potential improvements in the safety of operating power reactors and in focusing regulatory resources more effec-tively on the more risk-important issues and activities.
In addition, the NRC has a continuing responsibility to ensure that the risk to the public due to nuclear power plant operations is maintained at acceptable levels.
The NRC must therefore ascertain that licensees have in place (and maintain) adequate procedures for the installation, operation, maintenance, and test-ing of systems and equipment important to safety, based on a level of reliability that is commensurate with their influence on an overall plant risk.
4-4
The NRC must also publish procedures to ensure that plant operators are
[
i knowledgeable with regard to principal accident sequences and are trained to respond appropriately to abnormal events.
l The NRC is about to implement a severe accident policy statement.
Proce-i 4
i dural changes and modest equipment modifications may offer attractive alternatives to major plant modifications or other regulatory requirements that may be imposed in connection with closure of this issue.
There fore, I
there is a need for methods that will better quantify the risk-reduction j
effectiveness of such options to support mandated value-impact evaluations 1
of any proposed new requirements.
i l
4.2.2 Research Program Description 1
This program will develop the methods and supporting systems necessary to provide
[
(and update) information contained in existing (and future) PRAs into a readily accessible data base with output formats that will assist the licensing and inspection staffs in the conduct of their activities.
These activities include (1) organizing and setting priorities for inspection functions based on plant t
status, (2) establishing limiting conditions of operation, (3) establishing and revising plant technical specifications (including allowable outage times and surveillance test intervals), (4) responding to requests for exemptions from l
)
regulatory requirements, and (5) evaluating the risk impa::t or design changes.
Longer-range plans call for evaluating the feasibility of extending the methods j;
to plants that have not completed PRAs.
i This program will also develop, to the. extent practical and feasible, the manage-l ment structure, procedures, methods, and requirements necessary to ensure that nuclear power plant safety systems meet and maintain desired reliability levels.
The program will use information gained from nonnuclear applications of reli-j ability assurance techniques and will be developed so as not to duplicate either the requirements of Appendices A and B to 10 CFR Part 50 or other regulatory requirements related to reliability maintenance.
The research will produce i
guidance, suitable for use by NRR, IE, and industry, on the adaptation of con-s i
ventional reliability assurance program elements'within the institutional ar.d technical context of nuclear power plant operations.
(
l The major research products will be:
i 1.
Final report of the results of demonstration and recommendations for scope and content of effective reliability assurance program (1987).
I 2.
Based on risk assessment insights and updated analysis of dominant accident sequences, information from PRA analyses to be developed and formatted to t
aid in developing and setting priorities for IE inspection activities and i
licensing decisions (continuing program, 1987-1991).
i 3.
Report detailing recommendations on the application of generic information from the Accident Sequence Evaluation Program (ASEP) in resolving out-
{
standing safety issues at operating plants (1987).
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4-5 l
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4.2.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
The 1987 funding level has required a substantial realignment of research in this program element.
FY 1987 resources have been directed toward completing high priority programs to close the severe accident issue and to make available important risk-based information to the licensing and inspection staffs.
As a result, it has been necessary to defer and delay other important research.
This research would have provided more detailed technical support for the continued resolution of several outstanding reactor safety issues, for syste-matic examination of potential safety issues associated with advanced reactor design, for needed improvements in safeguards and transportation regulations, and for better methods to assess the risks of external events and common cause failures (e.g., fires) in operating reactors. The full impacts of funding limitations on this program and any additional programmatic changes will be reassessed after the FY 1987 appropriation legislation is completed.
4.3 Severe Accident Risk Assessment The Commission's policy statement on severe reactor accidents and standard plant designs set forth its intentions regarding future regulatory actions necessary to resolve safety issues relating to reactor accidents more severe than the design basis accidents used in current regulatory practice.
In this policy statement, the Commission set forth its position that, based on currently avail-able information, existing (licensed) plants pose no undue risk to the public health and safety and, further, that no immediate action on generic rulemaking or other regulatory changes is warranted at the present time.
However, the
. Commission also pointed out that, in many of the PRAs conducted to date, unique vulnerabilities to severe reactor accidents have been identified.
As a result, the Commission stated its intent to formulate an integrated, systematic approach to examine each nuclear plant now operating to identify plant-specific risk contributors that might be overlooked without a systematic search.
Summarized below are the principal elements of the risk assessment research being conducted to support the development of an integrated, systematic approach for closure of the severe accident issue.
It should be noted that a more complete description of the overall RES program on severe accidents is provided in Chapter 6.
4.3.1 Major Regulatory Needs and Their Justifications 1.
Identification and probabilistic evaluation of accident sequences, including uncertainties, that would have the potential for leading to severe core damage or core melt (1987-1989).
Justification:
Identification and description of the principal accident sequences that could lead to core damage will provide an essential element for the systematic analysis of the risk-reduction effectiveness of back-fitting safety systems on operating plants and for other NRC programs and applications, including incident reviews, assessment of generic issues, and real-time diagnosis and prognosis of incidents in progress.
This information is essential to the implementation of the severe accident and safety goal policies.
4-6
2.
Validation of the information base of accident sequence likelihood charac-teristics based on recent PRA studies, Licencee Event Report data, and updated plant information (1987-1989).
Justification:
Future decisions on the incorporation of additional safety requirements into the regulations, elimination of unnecessary requirements, and backfitting on operating plants require that the quantification of accident likelihoods be made using the most recent data available.
3.
Automation of risk assessment tools (event tree and fault tree models) and the accident sequence likelihood characteristics information base (1987-1989).
Justification:
Automation or computerization of the accident likelihood evaluation results can improve the NRC plant operational assessments for determining the risk status of plants from implication of generic issues, backfits, etc.; evaluating operational experiences to determine the effec-tiveness and pertinence of NRC regulations; ensuring NRC effectiveness during emergency response situations; and improving the effectiveness of NRC inspections.
4.
Development for regulatory use of a standardized, fast-running risk assess-ment code more readily understandable, amenable to modification, and includ-ing the capability to assess the uncertainty associated with risk estimates (1987-1988).
Justification:
The Commission's policy statement on severe reactor accidents (NUREG-1070) discusses NRC efforts concerning the need to consider severe accidents in the regulatory process.
The risk codes to be developed in this program will be the principal risk codes used in implementing the severe accident policy and in making related regulatory decisions.
Correc-tion of known deficiencies in these codes is necessary prior to their use in this context.
An evaluation of the uncertainties associated with the results of calculations made with these codes is also necessary to provide i
improved perspectives on the decisionmaking process using risk perspectives.
5.
Up-to-date analyses of the predicted probabilities and consequences (risk) from severe accidents, including improved estimates of the costs and effec-tiveness associated with adopting alternative designs, safety features, and operating procedures, applicable to the current generation of LWRs.
This includes development of strategies for identifying simple procedural changes or minor equipment modifications to reduce the risk associated with severe accidents in a cost-effective manner, to aid in the implemen-tation of severe accident policy and the resolution of outstanding safety issues (1987-1990).
Justification:
The consequence and risk reevaluations that will be con-ducted in this program will provide a baseline level of risk from which severe accident decisions can be made and risk-reduction and backfit analyses can be performed.
These assessments will require use of the best available information on accident frequencies, accident phenomenology (including results of updated estimates of the associated source terms),
and the potential effectiveness of various operating procedures for miti-gating the consequences of severe accidents.
4-7
4.3.2 Research Program Description The principal thrust of the severe accident risk research program has been directed toward the preparation of an integrated risk assessment report (NUREG-1150) that would profile and characterize the risk presented by selected operating reactors that typify basic vender and containment types.
The plants selected for the initial phase of the study are Surry, Peach Bottom, LaSalle, Grand Gulf, Sequoyah, and Zion (representing subatmospheric, Mark I, II, and III, ice condenser, and large, dry containments, respectively).
Such a report is necessary to support NRR's efforts to assess plant evaluation methodologies being developed by the Industry Degraded Core Rulemaking (IDCOR) program and to provide an improved technical base for implementing the Commission's severe accident policy.
This report will use source term calculation methods set forth in NUREG-0956, " Reassessment of the Technical Bases for Estimating Source Terms." This effort will continue into FY 1987, extending the reference plant analyses to a Babcock and Wilcox and a Combustion Engineering plant.
This program will also evaluate, as needed, residual risk-important issues identi-fled as a result of public comments on the draft report and new research results.
As presently conceived, this report is intended to present:
1.
An assessment of core melt frequency and risks associated with each of the reference plants.
2.
Perspectives on the elements driving risk-dominant accident sequences.
3.
A characterization of related uncertainties (qualitative and quantitative) in the assessments.
4.
Results of the evaluation of the effectiveness of several risk reduction features for the reference plants.
5.
Discussions of insights and perspectives gained from plant-specific studies, compared to insights gained from more generic plant models and data bases, such as existing in the ASEP.
6.
Recommendations for future regulatory use of reliability and risk information.
For the long term, it is expected that the data base developed in this program will be maintained, updated, and extended to plants that have not completed a formal PRA.
Where improvements are made in risk assessment methods (see Section 4.1) or where important new data are developed (see Section 6.1), these risk assessments will be updated or will be appropriate for meeting the then current needs of the licensing and inspection staff.
The major research products will be:
1.
Consequence and risk evaluations performed for selected reference reactor plants representing major vendor and containment types (1987-1989).
2.
HELCOR code documented for use in implementing the Commission's severe accident policy and the continued resolution of the outstanding safety issues (1987).
4-8
l 3.
Automation of risk assessment models and accident sequence likelihood characteristics information base (1987-1989).
4.
Report providing generic assessment of costs and risk-reduction potential of alternative safety features applicable to specific classes of LWRs (1987-1989).
5.
Evaluation of feasibility of using PRA to improve reliability of existing plant systems (1987-1991); determination of plant-specific severe accident management strategies (e.g., venting, depressurization) that are effective and of low cost in mitigating the consequences of core melt (1987-1990).
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5.
THERMAL-HYDRAULIC TRANSIENTS The purpose of the research is to produce the information necessary for the NRC to respond to new and technically complex regulatory issues affecting the design, operation, and maintenance of nuclear power plants.
In practice, this requires the ability to understand and describe the processes of heat transfer and fluid flow in light-water reactors for the range of plant transients that may occur.
This includes the design basis accidents required to be analyzed in license applications as well as other transients that actually occur in plants or may be postulated for the different plant designs in operation, under construction, or envisaged in the United States.
The principal products of thermal-hydraulic research are analytical tools (computer codes and their associated application methodology) to understand and predict the plant response to disturbances from normal operating conditions.
The codes model the plant behavior by describing the processes of heat transfer I
and fluid flow that occur in a plant.
The codes were developed through use of basic physical laws and information from thermal-hydraulic experiments.
Such experiments are conducted in integral and separate-effect facilities.
An integral facility is a scaled representation of a plant with all the major components present to provide information on overall system response and the interactions of different phenomena.
In contrast, a separate-effect facility studies a particular component or phenomenon in greater detail.
Each experi-mental facility has its own particular limitations associated with scaling and other design compromises that preclude direct extrapolation of experimental results to the full-scale plant.
Rather, the computer code provides the required link.
Furthermore, the codes allow different power plant transients to be studied without the need to embark on experimental programs for each separate case.
The code development and experimental programs have operated according to a feedback process. As different scenarios were encountered or postulated and
]
potential code deficiencies identified, particular experiments were run to obtain the data necessary to establish the code accuracy or to improve the code.
The interlinkage of code development and experimental programs is such that it is difficult for one to exist without the other.
The best-estimate analysis of transients using thermal-hydraulic computer codes provides a means of understanding plaret response.
In this way, the NRC is able to judge the design and operation of the plant and to develop appropriate rules and guides.
The proposed revision to Appendix K to 10 CFR Part 50 to permit more flexible and economic operation of plants is one example.
The pressurized thermal shock (PTS) rule, based on a program of experiments and computer code analyses, is a second example.
A third is the analysis of decay heat removal in Babcock and Wilcox (B&W) plants, which assisted in establishing the technical basis for continued operation following the recent Davis-Besse loss of-feedwater j
event. These are some examples of the approach taken by the research program, that of an interrelated experimental / analytical effort to obtain validated technical bases for decisionmaking.
5-1
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The research conducted since the adoption of S 50.46 of 10 CFR Part 50 and Appendix K in 1973 has resulted in a greatly improved understanding of the phenomena that may occur in a loss-of-coolant accident (LOCA).
A summary of this completed LOCA research is being prepared for publication in 1986. With the constraints currently imposed by Appendix K, a great deal of effort expended on LOCA analysis is based on assumptions that do not closely resemble the physical phenomena as they are now understood.
Moreover, distortions created by the use of artificial conservatisms can be a net detriment to the overall I
safety of plant design and operation.
Safety is best served through the use of best estimate analysis, with appropriate accounting of uncertainties.
Revision of S 50.46 and Appendix K is now under way to allow best-esticate rather than artificially conservative analysis.
The proposed rule will be issued for public comment in 1986.
This is expected to result in improved allocation of resources by the NRC and the nuclear industry.
The revised rule may allow relaxation of plant operating limits, reduced number of plant trips, core-loading. schemes for improved fuel utilization, reduced surveillance requirements, lower rates of embrittlement of pressure vessels, and operation at increased power within the limits of turbine and condenser capacity.
The economic benefit is expected to be significant, and both the NRC and the nuclear industry will be able to concentrate on more important safety issues.
5.1 Separate-Effect Experiments and Model Development This research consists of experiments and analytical work to provide data and models describing specific phenomena important to understanding and predicting reactor transients.
This work differs from integral experiments (see Section 5.2) in that important phenomena are investigated separately and in more detail than is possible in integral experiments, which simulate all aspects of the particular transient.
Examples of separate-effect experiments include detailed heat transfer studies and fluid mixing research to address PTS concerns.
The products of this research are specific models of phenomena used in the computer codes (see Section 5.3).
Separate-effect research is also used to address the scaling capability of the computer codes.
Most integral tests used to assess code capability are signifi-cantly smaller than a reactor system.
Thus, uncertainty exists as to the ability of the computer codes to predict transients in full-scale plants.
The system piping in an integral test is significantly smaller than the piping in a plant.
Separate-effect studies of two phase flow in pipes and other components are i
therefore necessary to ensure that the flow patterns in integral system tests are not significantly different from those that would occur in a plant.
Smaller-scale inte0ral tests using alternative scaling principles are also conducted to test the ability of the codes to predict results from facilities of various sizes and to test scaling theories.
Much of the separate effect research is performed at universities.
This provides a base of thermal-hydraulic experts and students to maintain this technical expertise. A solid community of technical experts in the field of thermal hydraulics is required if the NRC is to maintain the capability to address thermal-hydraulic issues in the future.
5-2
5.1.1 Major Regulatory Needs and Their Justifications 1.
Evaluation of integral transient behavior from facilities of varying size, to resolve scaling issues (1987).
Justification:
Results of scaled reactor system tests must be extended with confidence to full-sized reactors.
Small test facilities are needed to support major facilities to investigate the scaling of the phenomena.
Scaling also'must be understood to determine if small facilities can be used to replace larger and more expensive facilities as the identified work in the larger facilities is completed.
2.
Revision and assessment of heat transfer package in the RELAP and TRAC codes, to be used in evaluating operator guidelines (1987).
Justification: The heat transferred from fuel rods to water is calculated in current RELAP and TRAC code versions according to correlations such as the "re-wet criterion" and critical heat flux (CHF) that are known to be conservative when used in emergency core cooling system (ECCS) licensing evaluations.
The conservatisms have been found to be unrepresentative of actual phenomena.
More representative correlations should therefore be developed and introduced into both codes (RELAP and TRAC) in order to reduce the conservatism and uncertainties in the calculated results.
3.
Validation of the effects of tube rupture and iodine transport through the secondary system for the balance of the plant (80P) in advanced codes for use in evaluating anticipated transients (1987).
Justification:
NRR routinely uses the TRAC and RELAP computer codes to evaluate various accident scenarios involving steam generators during the licensing process.
Because current steam generator models in TRAC and RELAP are not based on directly relevant experimental data, unknown uncertainties may be present in the above analysis.
The data base gene-rated by these experiments will quantify these uncertainties and point to the need for code modification, if any.
4.
Fluid temperature fluctuations at the intersection of the high pressure lnjection (HPI) lines and the cold legs of PWR systems at a variety of HPI and cold-leg fluid mixing conditions (1988).
Justification:
Failure of the weld that joins the HPI line and cold leg woulu simultaneously initiate a small-break LOCA and reduce the ability of the ECCS to mitigate such an event.
Only a limited amount of experimental data is available on thermal fluid mixing at the HPI and cold-leg inter-face.
This information was requested by NRR to determine the possibility of cyclic fatigue failure at these welds.
5.
Boron mixing in the lower plenum of BWRs during a postulated anticipated transient without scram (ATWS) (1987).
Justification:
Data on mixing of cold boron solution are required so that computer codes can accurately analyze ATWS in BWRs.
6.
Tests of auxiliary feedwater spray flow and heat transfer characteristics in a full-scale separate effect facility modeling a sector of B&W once-through steam generator (consisting of at least three tube support plates and suitable array of steam generator tubes) (1991 and beyond).
5-3
Justification:
In B&W-designed once-through steam generators, auxiliary feedwater is sprayed onto the steam generator tubes in the upper region of the boiler section.
The ability of this spray to provide primary system cooldown is strongly affected by its ability to penetrate the tube bundle and the tube support plates.
These data have a great impact on transient responses, e.g., thermal shock, and no transient or steady state data exist for large bundle geometry.
5.1. 2 Research Program Description A small-scale B&W reactor simulation loop at the University of Maryland has been completed to provide (1) separate-effect testing of important transients, (2) scoping tests for input to the B&W Multiloop Integral System Test (MIST) facility, and (3) data to investigate the scaling of MIST results to a full-scale plant.. In addition, separate-effect testing of the B&W hot leg is being performed to determine conditions responsible for flow interruption and resumption and to evaluate the applicability of the results from MIST, which has a significantly smaller hot leg than that found in a plant.
Development of two phase flow models and correlations is continuing for the purpose of providing a foundation needed for validated LWR safety analyses.
This program will be investigating (1) inverted annular flow and the mechanism for jet breakup, (2) similarity laws under natural circulation conditions, and (3) a hydrodynamic model for entrainment of water from a pool.
These data will be used to improve flow models in the codes as required.
A thermal fluid mixing program is being planned to obtain thermal-hydraulic data that can be used to develop and assess models that describe the extent of thermal fluid mixing in a reactor downcomer and cold leg as a result of ECC injection and to apply to boron mixing in BWRs.
This research will provide NRR with valuable support in evaluating the PTS issue and ATWS analysis.
Data from the completed MB-2 program is being used to study steam generator heat transfer during small breaks, tube rupture events, and other system tran-sients.
Effects of tube rupture, including iodine transport through the secon-dary system, are also being investigated under this and other programs at Northwestern University, Massachusetts Institute of Technology, and the Oak Ridge National Laboratory (ORNL).
Scoping studies will be perforced in 1987 in an air-water visual scale model of a sector of a once-through steam generator (OTSG) with an auxiliary feed-water spray.
The facility, constructed of plexiglass, will consist of 200 10-foot-long tubes and three tube support plates. Observations and data will aid in the design of a steam-water and Freon sector facility and in understand-ing OTSG response characteristics.
A low pressure (250 psia) sector facility will be constructed in 1989 and 1990 in which reduced pressure steam-water tests and full-scale pressure Freon tests of auxiliary spray fluid flow and heat transfer will be performed in 1990 and 1991.
The proposed facility will consist of a large array of full-scale steam generator tubes of reduced length and seven tube separator plates.
Frc= these tests, improved fluid flow and heat transfer models will be developed for TRAC and RELAP.
5-4
The major research products will be:
1.
Evaluation of scaling of integral test data and of scaling methods at
{
alternative reactor system test facilities (1987),
2.
Revision of heat transfer packages in RELAP5 and TRAC, based on assessment results (1987).
3.
Improved models for iodine transport through the secondary side (1987).
)
4.
Thermal mixing model to predict fluid temperature fluctuations at the HPI i
and cold-leg interface (1988).
5.
Boron mixing model (1987).
6.
OTSG auxiliary feedwater spray thermal-hydraulic models (1991 and beyond).
5.1. 3 Deferred /Unfunded Research Needs (as of February 1, 1986)
In order to provide a stable base of expertise in fundamental understanding of thermal-hydraulic processes under reactor accident conditions, a strong research i
program is needed in universities.
The development of university " centers of excellence" for both analyses and experiments was severely curtailed by FY 1986 and FY 1987 budget restrictions.
Past experience has shown that some of the best work on fundamental understanding of reactor safety thermal-hydraulic processcs has come from universities.
Construction of visual loops at univer-sities to study scaling will not be funded.
5.2 Integral Systems Experiment _s The NRC has used integral thermal-hydraulic facilities to investigate reactor thermal-hydraulic transients since the early 1970s, starting with an early version of Semiscale.
Integral facilities are scale models of reactor systems that simulate, to varying degrees of completeness, the entire reactor system and the entire transient of interest.
Integral facilities differ in this regard from separate effects or model development tests, which focus in more detail on a specific component of the reactor system or attempt to isolate certain phenomena for detailed study.
Early work in Semiscale, LOFT, and the Two-Loop Test Apparatus (TLTA), as well as recent work in the 2D/3D facilities, i.e., Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF),
and Upper Plenum Test Facility (UPTF), concentrated primarily on simulating large-break LOCA transients to investigate the performance of the ECCS during this transient.
Later upgrades of these facilities and new facilities, such as Full Integral Simulation Test (FIST), Once-Through Integral System (OTIS),
MIST, and ROSA IV, have been used primarily to investigate small-break LOCAs and other nonbreak transients.
i The results of these tests have been used to support the licensing of nuclear reactors in two ways:
(1) to infer how a plant would perform during a similar transient and (2) to assess computer codes used to calculate reactor transients.
t The latter use (i.e., code assessment) is generally regarded as the proper use of integral test data since limitations in facility scaling or representation r
5-5
i of the specific plant or transient can make the direct use of integral test results misleading.
The end product of the thermal-hydraulic transients program is assessed computer codes that can be used with confidence to calcu-late reactor transients.
Integral test data form a major data base for use in assessing these computer codes and in evaluating the uncertainty of these codes when calculating plant behavior.
Early integral testing concentrated on large-break LOCAs and was directed primarily toward the development and assessment of best-estimate computer codes used to calculate large-break LOCAs.
Licensing application was through the use of the best estimate codes to assess the conservatism of licensee evaluation model codes.
In recent years, however, the use of best-estimate computer codes in the licensing process, supported by integral testing, has significantly increased.
Following the TMI-2 accident, NRR required that small-break LOCA evaluation models approved for use in licensing be assessed against integral test data.
Tests conducted in FIST, Semiscale, and LOFT were used by the vendors to assess these models.
In addition, a new program, the Integral System Test (IST) program, was developed to provide applicable data with which to evaluate the B&W small-break LOCA models. This also marked a major shift in emphasis on the part of the NRC from concentrating on the large-break LOCA to studying more probable transients.
A corresponding shift toward using more realistic analyses to determine the actual effects of plant equipment and procedures resulted in the increasing use of the NRC-developed computer codes and integral test data in the licensing process.
One example of this occurred shortly after the TMI-2 accident when the issue was raised whether to trip the main reactor coolant pumps or leave them running.
Integral tests in both Semiscale and LOFT were conducted to support the results being provided by computer calculations.
A number of other issues have been raised requiring dual code / experimental analy-ses, ranging from generic issues such as PTS to more specific concerns such as the capability for feed and bleed cooling in the Davis-Besse plant.
In addition to using codes and integral test data to evaluate the small-break LOCA and the other issues discussed above, the large-break LOCA is now being revisited in terms of more realistic analyses supported by the integral test data.
General Electric (GE) has recently had a more realistic large-break LOCA evaluation model, SAFER, approved for licensing jet pump BWRs and has submitted a similar request for the review of a more realistic evaluation model for non-jet pump BWRs.
In addition, the proposed revision to S 50.46 and Appendix K i
to 10 CFR Part 50 would allow realistic ECCS evaluation models.
The GE evalua-tion model was directly supported by integral test data from TLTA and FIST, and any future use of the proposed revision to S 50.46 would require similar support by integral test data.
Thus, it is expected that integral test data will be used more and more directly in licensing.
The computer codes are now mature and all identified integral test data needs for initial code assessment have either been satisfied or programs are in place to obtain these identified data.
We do not, however, believe that the need for integral test data will end after the identified programs have been completed.
Rather, we see a continuing need to provide an integral test capability.
Reac-tor events such as the Davis-Besse transient or issues such as " pumps on or of f" have occurred at a rate of at least one per year, and it is prudent to 5-6
l expect this to continue until experience proves otherwise.
In each case, NRR has requested analyses to resolve the particular safety issue and integral tests to back up the analyses because the event or issue has been sufficiently different from previously calculated events that experimental confirmation of the code applicability is desired.
Thus, we see that the need for a capability to provide integral test data will continue beyond the currently identified programs.
Integral facility testing is very closely linked to the transient models and codes program.
Computer calculations of plant transients help to define the issues and uncertainties that need to be addressed experimentally either through integral or separate-effect testing.
Computer calculations are also used to help design integral test facilities, determine required instrumenta-tion ranges, identify facility atypicalities, and plan testi.
The interrela-tionship between the codes and facilities is iterative, with code results used to design the most useful tests and the results of the tests used to assess or improve the code predictions as necessary.
All the integral facility projects include significant code calculation tasks.
Integral testing is also interrelated with the separate-effect program.
Separate-effect tests investicate certain phenomena or reactor components in more detail than is possible in integral tests, which attempt to investigate the entire transient.
Integral and separate-effect tests represent, therefore, two separate and distinct inputs to the assessment or improvement of the computer codes.
In certain instances, however, separate effect tests are used directly in support of both the codes and the integral facilities.
Separate-effect tests are used to investigate the effects of possible atypicalities of the integral facilities and to either suggest modifications to the facilities or develop cautions to be used when extrapolating integral facility data to a plant.
5.2.1 Major Regulatory Needs and Their Justifications 1.
Experimental data from scaled simulations of reactor transients, including steam line breaks, feedwater line breaks, small-break LOCA without HPI, large-break LOCA with upper plenum injection (UPI), and scaling tests in Westinghouse and Combustion Engineering geometry (1987).
2.
Experimental data from scaled simulations of small-break LOCA and other reactor transients in B&W geometry as a result of TMI-accident concerns (1987).
Justifications of Above Needs:
1.
A need exists to perform calculations of transient conditions in order to audit the acceptability of operating precedures.
The actual response of the systems, as well as the indicated information provided by various instruments during transients, must also be evaluated to ensure that the operator can correctly identify the transients and take appropriate actions.
It is especially important that operators understand the way in which their machine operates and that their training and operating procedures are such that they are able to head off potential accidents and mitigate potentially serious accidents.
These tests will be used directly and, 5-7
through use of the data, to assess computer codes, to identify precursor events, and to improve our understanding of PWR and BWR transients and (in conjunction with the work on operator interactions) the adequacy of operator guidelines.
2.
Revisions to operator guidelines, safety system setpoints, and additional safety systems are periodically proposed to the NRC by reactor owners.
The data obtained from these test programs, in conjunction with improved calculational capability, will allow a better evaluation of the adequacy of procedures and safety systems and the influence of changes on safety.
3.
Past LOCA research has identified a large margin in the LOCA evaluation model calculations, which indicates the potential for relaxing some operating restrictions through the use of improved evaluation models.
This research will provide additional information for use by the NRC staff in proposing revisions to evaluation model requirements and for use during review of improved evaluation models submitted by applicants.
5.2.2 Research Program Description The Semiscale facility completed planned testing in 1986 and was immediately shut down.
The facility is now being preserved in a cold shutdown state.
Documentation of test results will continue into 1987.
thder a bilateral agreement with Japan, data from ROSA IV will be available.
Yests of small breaks, natural circulation, steam generator heat transfer, alternative ECC systems, and transient recovery techniques will be conducted in 1986 and 1987.
This program will provide integral PWR data beyond the planned phasecut of Semiscale and data from a new and larger facility for further assessment of the scaling capability of RELAP5 and TRAC.
The MIST program has recently been formulated to investigate the unique fea-tures of the B&W reactor system.
Experimental data are needed to address licensing concerns, to verify operator guideline procedures, and to assess code capabilities to predict B&W design related phenomena.
The MIST facility will employ two hot legs, and four cold legs (2x4), two steam generators, and pumps to model the lowered-loop B&W design.
Initial MIST testing started in 1986 and will include 6 months of debug and characterization tests followed by 6 months of composite testing.
The composite test matrix consists of 41 tests divided into four general groups-small-break LOCA, natural circulation, steam generator tube rupture, and feed-and-bleed cooling.
Follow-on testing in the MIST facility is currently planned for 1987 and 1988 and is referred to as the MIST-IV program.
Tests will be conducted at decay power level, as in the current program, but will require a configuration change to permit the simulation of non-LOCA transients.
Fewer than 20 tests will be performed to provide a minimally acceptable data set on steam generator behavior as a result of feedline break, loss of feedwater, and overfill of the steam generator; risk-dominant transients such as small-break LOCA without HPI and station blackout; and alternative recovery procedures.
Analyses of the test results would be completed by 1989.
5-8
I i
However, additional testing in MIST with 100 percent scaled power is necessary to properly address some transients such as feedline break and steam generator overfill.
The estimated cost of performing full power tests in MIST is several million dollars.
The NRC will therefore request EPRI, B&W, and the B&W Owners' Group to share the funding of these full power tests as part of a continuing i
cooperative effort to provide an adequate data base for evaluating code capabil-5 ities.
It is anticipated that low power and full power testing in MIST could be completed by 1990.
i The Department of Energy (DOE) owns and operates the LOFT facility, which is under the sponsorship of the Organisation for Economic Co-operation and Develop-j ment (0 ECD) LOFT project, and funding is shared by approximately 10 countries and agencies. The NRC is a member of the project based on a one-time contri-i, bution of $25 million in FY 1983.
Included in the test program were a complete loss of feedwater; small hot-leg breaks with and without primary pump opera-l tion; small cold-leg break without HPI and delayed auxiliary feedwater; double-ended large breaks in the cold leg; fission product release, transport, and i
deposition during large-break LOCA; and fission product release, transport, and deposition during the first phase of a severe accident (similar to TMI-2 j
conditions).
This test program was completed.in 1985, with follow-on analyses extending through 1987.
Identified testing has been completed in the FIST facility, but the facility will be maintained so as to be available if the need for additional tu ting is identified.
i
~Full-scale testing of ECC performance in the international 20/3D facilities 1
(UPTF and SCTF) will be accomplished during 1986 and 1987.
Data will be
~
provided on fluid-fluid mixing (PTS related), ECC bypass, upper plenum deentrainment, fallback, steam binding, flow blockage, and UPI.
I i
RES has initiated a study of NRC's long-term needs for integral facilities and how best to meet these needs. The study includes evaluation of retaining one i
or more existing facilities versus constructing replacement facilities using alternative scaling methods to permit the most cost-effective maintenance of test facility capability as well as to provide additional code assessment data.
i The decisions on these longer-term experimental needs will be heavily influenced by the management-level discussion on the objectives of the code work.
Based on preliminary results, the long-range plan assumes that the design work on an I
advanced integral facility will be performed in 1987 through 1989.
Necessary site modifications at the Idaho National Engineering Laboratory and facility construction would take place from 1989 through 1991.
The current choice of 1
facility is a 1/10 linear-scale model of a B&W plant, i.e., one that is generally a 1/10 photographic reduction of the plant.
This facility would y!
provide multidimensional responses and interactions during transient simula-tions that have never before been performed and could significantly alter our 1
understanding of the response of reactors being modeled.
The major research products will be:
1.
a.
Semiscale data to be used to assess small-break LOCA without HPI (1987).
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b.
Final topical reports for Semiscale test series, including (1) small-break LOCA without HPI, (2) liquid holdup, (3) power loss, and (4) steam generator tube rupture (1987).
c.
Full-scale fluid mixing and ECC bypass data from UPTF and SCTF (1987).
d.
Data on effectiveness of UPI and summary report (1987).
e.
Summary report on refill and reflood testing in CCTF and SCTF (1987).
f.
Assessment of TRAC-PF1 capability using ROSA IV test data (1987).
2.
a.
Experimental once-through-steam generator data analysis report from MIST (1987).
b.
Assessment of TRAC-PF1 and RELAP5 using MIST data (1987).
5.2.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
Testing in Semiscale or other facilities to assess the ability of codes to analyze the response of plants equipped with upper-head injection to large-break LOCAs, as requested by NRR, has not been included in these plans because of its expected removal as a safety injection system from these plants.
Because of FY 1987 funding deficiencies, there will be no United States integral l
facility to perform tests on Westinghouse, Combustion Engineering, and General l
Electric geometry after FY 1986.
Possible restoration of this capability will be deferred until FY 1991 or beyond.
5.3 Code Assessment and Application The NRC staff requires the ability to independently review plant designs and safety analyses submitted by license applicants and to review operating experi-ence.
This requires foremost that a sound physical understanding be obtained of the important phenomena that may occur during transients in operating power plants.
In practice, this requires the ability to understand and describe I
through the use of codes the processes of heat transfer and fluid flow for the range of plant transients that may occur.
The codes were developed through the use of basic physical laws and information from thermal-hydraulic experiments conducted in integral and separate-effect l
l facilities.
Each facility has its own particular limitations associated with l
scaling and other design compromises that preclude direct extrapolation of I
experimental results to the full-scale plant.
Rather, the codes provide the required link.
Furthermore, the codes allow different plant transients to be studied without the need to embark on experimental programs for each separate I
case.
The thermal-hydraulic codes developed, i.e., TRAC-PWR, RELAPS, COBRA, RAMONA, and TRAC-BWR, provide best-estimate predictions of plant responses to transients j
for use by NRR.
Their development was focused initially on the analysis of l
large-break LOCAs.
This emphasis was changed, as a result of the Reactor Safety i
5-10
Study (WASH-1400) and the TMI accident, to small breaks and transients.
The focus on transients extended the range of events that the codes must be capable of analyzing and increased the requirements for code accuracy in modeling plant behavior.
The development of two PWR codes was supported since 1974, prompted by the strong need that existed for the analysis of transients and breaks.
Both TRAC and RELAP had capabilities and limitations that precluded concentrating strictly on one or the other.
Both codes have now reached the level of being relatively capable, versatile codes, and no major new versions are planned.
Both will be maintained since they exist and each is valuable.
However, future resources for model improvement and assessment will be concentrated on TRAC, and only selective model improvements to RELAPS will be made on a case-by-case basis.
Error corrections will be made on both codes.
User-convenience features will be added to both codes on a case-by-case basis.
The COBRA code was also developed for PWR analysis.
This code has the capabil-ity for detailed subchannel analysis and is not a systems cooa in the same sense as TRAC although it was linked to TRAC-P02 to produce COBRA / TRAC.
The RAMONA code was developed initially in Sweden and obtained by the NRC in 1978.
Further development of the code was performed at Brookhaven National Laboratory.
The code is used for special applications that require three-dimensional calculation of core power and neutronics, in particular, for reactivity transients and ATWS or partial ATWS events in BWRs.
Independent assessment is being addressed as part of an international program.
For a number of years, different countries and organizations have been using earlier versions of TRAC-BWR, TRAC-PWR, and RELAP and communicating their user experience to the code developers on an informal basis.
During 1984 and 1985, the NRC organized and planned a more formal effort devoted to the assessment of the most recent frozen versions of the codes, i.e., TRAC-PF1/ MOD 1, RELAP5/M002, and TRAC-BD1/M001.
The basis of the program is a series of bilateral exchange agreements concluded or nearly concluded with Austria, Belgium, Finland, France, Federal Republic of Germany, Italy, Japan, Korea, Netherlands, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the Commission of the European Communities.
The duration of the program is approximately 5 years, from 1985 to 1990. The participants will perform assessments using data from their experimental facilities and power plants and provide the results to the NRC.
The NRC will use this information and the results of domestic assessment and applications of the codes to arrive at overall conclusions on the accuracy and reliability of the codes for best-estimate predictions of plant behavior.
This includes a quantification of code uncertainty.
The size of the uncertainty shows how accurately codes can predict transients.
In addition, the uncertainty for TRAC will be used in deciding when model improvement for TRAC should be considered to be completed.
The program is being coordinated through an inter-national program group, which is providing peer review of the codes and their assessment.
Establishment of the methodology for code uncertainty (accuracy) and its application will serve to quantify the applicability of the code assess-ment efforts.
5-11
The current focus for code assessment and applications includes:
Analysis of transients in operating power plants (e.g., the recent Davis-Besse event).
The NRC requires the capability to quickly analyze transients using computer codes and to perform experiments to investigate relevant phenomena in a well-instru-mented experimental facility.
Study of small-break LOCAs in B&W reactors.
Much research has been directed at Westinghouse and Combustion Engineering designs, and much of this is applicable to B&W plants.
There are, however, two principal features of B&W plants that require specific experi-ments aimed at their study:
(1) vent valves between the vessel upper plenum and downcomer and (2) once-through steam generators and auxiliary feedwater spray.
The experimental data are being provided through the IST program, which includes MIST.
The analysis includes use of TRAC-PWR and RELAP5.
Secondary system transients and breaks.
Most research has been oriented toward primary system transients and breaks.
Additional experiments and analyses are in progress or planned for additional study of secondary system effects.
This includes the ROSA IV program in Japan, BETSY in France, and SPES in Italy and also analyses using TRAC-PWR and RELAP5.
5.3.1 Major Regulatory Needs and Their Justifications 1.
Maintenance and improvement of TRAC and RELAPS codes (1987-1991).
Justification:
These codes are important to the licensing staff in per-forming system transient analyses to investigate various recovery tech-niques, understand phenomena that occur during transients, and review analyses submitted by vendors and utilities.
Model improvements in codes are to be limited to those that would improve accuracy and affect percep-tion of transients and accidents.
User requested features such as auto-matic initialization will be added.
2.
Assessment of computer codes for use in resolving licensing and safety issues (1990).
Justification:
Use of these codes to analyze highly complex reactor transients and accidents in full-scale geometry requires extensive assess-ment against test and plant data to ensure the reliability of results.
Regulatory decisions need to be based on broad and accurate technical information.
Assessment of the accuracy of the predictions is important to the NRC staff.
5.3.2 Research Program Description The strategy for the research in this element has been to develop methods and codes and simultaneously test them against data.
At an early date, each code is released to groups independent from the DOE laboratory developers (usually other DOE laboratories) for independent evaluation.
Since 1984, foreign organi-zations have been participating in the assessment effort.
5-12 l-
Collection and analysis of nuclear data and plant design and operating histories are ongoing efforts that support this element.
Technical information developed jointly with outside groups (EPRI, foreign governments) is often used to improve and assess the NRC computer codes.
RELAP5 and TRAC will be used in a fixed form in the 1986 and 1987 period.
This research program will develop code packages for analyzing complete LWR systems for transients ranging from anticipated transients through design basis events.
In addition, methods to predict system behavior for small-break tran-sients coupled with operator actions will be developed both for a fast-running and a detailed analysis code.
Methodology of code application to the solution of transients and accidents will also be established.
Physical models and numerical methods for prediction of two phase flow behavior will be improved in the codes.
Codes will be improved, and their accuracy assessed, to calculate operating reactor events and the front end of severe accidents.
Since 1984, the advanced system transient codes have been maintained and improved based on independent assessment results from the International Code Assessment Program (ICAP) and test data to be obtained from facilities such as UPTF, CCTF, SCTF, ROSA IV, MIST, and the new integral test facility (Section 5.2.2).
The major research products will be:
1.
a.
Development of a method for statistical uncertainty for large thermal-hydraulic analysis codes (1987).
b.
Reports summarizing research results on small-break LOCAs and opera-tional transients to assess their integration to regulatory process (1990).
c.
Improved versions of the advanced multidimensional two-fluid tran-sient analysis codes:
TRAC-PF1/ MOD 2 (1987), RELAP5/ MOD 3 (1987),
TRAC-PF1/ MOD 3 (1990), ard RELAP5/M004 (1990) (new versions to contain error corrections and user-convenience features; TRAC codes to also contain model improvements).
d.
Assessment of TRAC-BF1 by DOE laboratories (1987).
e.
Determination of code uncertainty (accuracy) for TRAC-PWR, TRAC-BWR, and RELAP5 codes (1990).
2.
Status of assessment results from ICAP and their significance in resolving licensing issues and improving reactor operations (1990).
5.4 Nuclear Plant Analyzer and Data Bank' The nuclear plant analyzer (NPA) program addresses the NRC need to be able to perform rapid analysis of transients in operating power plants.
The second main purpose of the NPA program is to reduce the costs to the NRC of performing analyses by improving the productivity of analysts and by increasing the calcu-lational speed to reduce computational costs.
The codes TRAC-PWR, RELAP5, and TRAC-BWR are necessarily quite complex and from the time a calculation is begun 5-13
.m
to the time the analysis is completed is a relatively long period.
The prepara-tion of input is time consuming and requires considerable experience and exper-tise.
The interpretation of output requires similar resources.
These factors combine to limit the use of the codes to a relatively limited set of experts located at Idaho National Engineering Laboratory and Los Alamos National Laboratory.
Part of the NPA effort includes the nuclear plant data bank (NPDB),
which is a means of storing and retrieving input models for United States power plants.
A fast running NPA for BWR-4s has already been completed at the Brookhaven National Laboratory and is available for use by NRR and IE.
The specific NPA objectives are to:
1.
Reduce the man-hours required to prepare input decks and obtain steady-state initialization through the NPDB part of the NPA.
The NPDB also provides for storage and retrieval of plant geometric information so that validated input models can be quickly obtained.
2.
Permit easy renodalization of the input model to determine noding sensi-tivity and to obtain the appropriate level of modeling detail for the transient being studied.
3.
Provide interactive capability simulating operator control of pumps and valves to allow calculation of the transient to proceed following different scenarios as the transient progresses, rather than a single, predetermined scenario.
This permits the evaluation of the effect of different operator actions or equipment malfunctions.
4.
Speed up existing TRAC-PWR and TRAC-BWR codes to reduce computation costs and to allow real-time analyses for plant simulations.
5.
Provide for colorgraphic information displays showing the state of the coolant throughout the primary or secondary system.
The colorgraphic presentation of data replaces numerous time-history plots for each node, which normally must be studied separately to understand and assimilate the results.
The productivity of the analyst is thereby greatly increased.
- 5. 4.1 Major Regulatory Needs and Their Justifications 1.
User-convenient system analysis codes for use in evaluating transients and accidents with the capability for interactive operational manipulations at midpoints during long transients (1987).
Justification:
NRC and contractor personnel need the capability to per-form analyses of full-scale LWRs in order to help resolve safety and licensing issues in a timely fashion.
These analyses should be easily initiated, should allow user interaction, and should provide easily under-standable output results in order to have high utility.
To meet these needs, the plant analyzer is being developed.
Initial versions are now in use by the licensing staff.
5-14
i 2.
Geometric and operating data for selected licensed plants to allow plant-l specific calculations to be performed (1987).
i Justification: The conduct of these analyses will be facilitated if the plant-specific data are contained in a data bank.
There must be a complete set of geometric as-built data and thermal-hydraulic and neutronic char-
- acteristics available in computer language.
These data will be automati-cally converted into input docks for the plant analyzer.
A program for obtaining and inputting plant-specific data will be developed for timely implementation.
4 5.4.2 Research Program Description As the development phase of NRC codes is nearing completion, more emphasis is being placed on making them available in a user-oriented form. This develop-ment is focused in two areas:
(1) display of the computed transient on terminal consoles so that the user can easily understand the calculated results and inter-act with the calculation, if desired;'and (2) incorporation of LWR plant data into a data bank that is easily accessible for the development of input decks for computer codes and the plant analyzer.
Colorgraphic replay of previous analyses will be available, as will selected experimental results.
In addition to user convenience, the plant analyzer and data bank will provide long-term i-benefits in the traceability and auditability of reactor data, code input decks, l
and reproducibility of results.
I The specifications of the plant analyzer reflect our current experience with the RELAPS and TRAC codes.
Since it commonly takes 4 to 6 months to prepare an input deck for such codes, it is planned to use information stored in a plant data bank to compile input decks in an automated fashion with minimal input from the user.
The data bank itself is in existence but the software contains i
errors.
A program for error corrections is in place.
Similarly, there is a need for interactive features to allow the user to follow a computed transient in detail and to intervene at some step to make a change in plant condition to replicate a projected action.
Current practice would require successive restarts of runs, approximately three per change for long runs, to follow a transient.
An interactive feature being developed will significantly reduce computational i
effort and will speed the acquisition of results.
Special plant analyzer versions j
of the advanced systems codes (Section 5.3) will be maintained.
The major research products will be:
1.
a.
Demonstration of construction of input decks for a 4-loop Westinghouse plant for a LOCA analysis using the NPDB (1987).
b.
Demonstration of typical PWR plant analyzer with incorporated NPDB (1988).
2.
a.
Typical PWR secondary systems and control systems implemented in NPDB (1988).
b.
Use of NPA and NPDB to analyze transients in full-scale PWRs to resolve licensing and safety issues (1987-1990).
P 5-15
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c.
Demonstration of typical BWR plant analyzer with incorporated NPDB (1990).
5.4.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
Because of difficulties in implementing the NPDB software and in obtaining plant data from utilities, the effective use of the data bank will have to be delayed. We expect that errors in the NPDB software for Westinghouse (4-loop plants) will be corrected by the middle of 1987.
The data bank capability will be expanded to B&W plants by the end of 1988.
Because of the FY 1987 budget cut, the advanced version of the nuclear plant analyzer, using parallel processors, was cancelled.
The advanced nuclear plant analyzer would perform computations ten times faster than the TRAC-PF1/ MODI code.
5.5 Technical Integration Center The emphasis of the NRC role relative to nuclear power plants has shifted in recent years from design basis evaluation of proposed new nuclear power plants to evaluation of the operational safety of operating plants.
The accident at TMI-2 focused attention on the need for this shift in attention and for emphasis on disciplines such as risk perspective, equipment reliability, and human factors.
In addition, transients in operating reactors continue to occur each year, including those that are judged " risk significant." There is every reason to believe that the frequency and risk significance of these events will continue in view of the large number of plants in operation and their increasing ages.
The staff's assessment of these events in operating reactors is greatly compli-cated by the wide variations in plant designs and operations, which could significantly alter the course and consequences of a transient, even for similar reactor designs from one vendor.
NRC's ability to determine the implications of actual, as well as postulated, events in operating reactors requires contin-uing, on-call access to an expertise base.
Because of the complexity of likely issues in the future, this base needs to consist of the best validated analy-tical methods available, experimental facilities to provide confidence in analy-sis results that have not previously been adequately validated, and trained personnel who are capable of applying the analysis methods and operating the experimental facilities and are familiar with nuclear plant operating experience.
5.5.1 Major Regulatory Needs and Their Justifications 1.
Ability to rapidly respond to postulated or actual events in operating reactors that have potential safety significance by performing analyses and experiments as needed to evaluate the safety implications for all reactors and to aid in formulation of appropriate regulatory actions.
Justification:
In the past, this response capability has been required, e.g., the Davis-Besse loss of feedwater on June 9, 1985.
It is antici-pated, in view of the number of plants operating in the United States and their increasing age, that similar events requiring regulatory staff evaluation will continue into the foreseeable future.
4 5-16 i
i 2.
Thorough assessment of the applicability of the models contained within major thermal-hydraulic codes, as well as the codes themselves, to be established for transients of interest to NRC for each type of operating reactor.
Justification:
The staff needs to understand the capabilities and limita-tions of NRC's thermal-hydraulic codes for their independent assessment of plant responses to postulated transients.
In addition, the staff should understand these capabilities and limitations for the constituent models in these codes in order to evaluate vendor codes using these same models.
3.
Capability to assess the relative ability of proposed design and opera-tions changes to improve plant risk and safety, in particular, an inte-grated analysis evaluation capability needed for B&W plants (1987).
Justification:
The staff's decisionmaking ability should be improved if key elements to decisionmaking related to plant transient response, particularly thermal-hydraulics, risk assessment, human reliability, and equipment reliability, are resolved.
Such an integrated method should be provided to the staff for independent evaluation of recommendations by the B&W Owners' Group to reduce the frequency and risk of B&W plant transients.
4.
Technical assistance to the NRC staff in assessing all the research results available, assessing their applicability and sufficiency for resolving current and postulated future safety issues, and formulating future research projects that will adequately satisfy the agency's research needs.
Justification:
The knowledge and expertise developed within the Technical Integration Center (TIC) is a valuable resource to aid the NRC staff in formulating research programs that are technically sound and can provide the needed results.
5.5.2 Research Program Description By agreement between the NRC and the TIC contractor, Idaho National Engineering Laboratory (INEL), the staff on all RES thermal-hydraulic research work at INEL, and to a limited degree those working on risk analysis and human reliability research, are available for staffing needed for timely resolution of priority issues.
The nature and timing of these priority issues cannot be defined in advance but would be defined as appropriate by the Regulatory Research Review Group, one member of which is from the TIC staff.
The TIC also provides manage-ment and planning staff for coordination, program review, planning, and prepara-tion of proposals for priority studies.
In addition, the TIC will establish the applicability of existing best-estimate thermal-hydraulic codes to the accident scenarios of interest to the NRC for each major nuclear power plant design variation.
The major research products will be:
1.
Reports, staff briefings, and other technology transfer, as appropriate, for staff resolution of priority issues (as required).
2.
Assessment of TRAC and RELAP applicability to B&W plants (1987), Westing-house ice-condenser plants (1988), and GE plants (1989).
5-17 n
3.
Integrated thermal-hydraulic, risk analysis, and human reliability methods available for staff assessments of B&W Oconee plant (1987) and others as needed.
4.
Update of thermal-hydraulic research plan (annually).
I a
i G
5-18 x
6.
SEVERE ACCIDENTS This program provides the data base and validated methodology for reassessing the regulatory treatment of severe accidents.
It includes the coordinated phenomenological research programs needed to develop a sound technical basis for NRC decisions concerning the ability of reactors to cope with these acci-dents.
The following elements are included in this chapter:
severe accident sequence analysis (SASA), in-vessel core-melt progression, h drogen generation 3
and control, fuel-structure interaction, containment analysis, fission product release, behavior, and transport, containment failure mode, fission product control, and severe accident risk and risk-reduction analysis.
The program outlined in this chapter anticipates that a significant level of confirmatory work will be required following the Commission's final decision on severe accident policy.
It also attempts to address areas of phenomenol-ogical uncertainty that were identified as part of the American Physical Society review of NRC source term research.
This program was also addressed in Revision 1 to NUREG-0900, " Nuclear Power Plant Severe Accident Research Plan."
6.1 Severe Accident Sequence Analysis Research uses the analytical assessment of plant accidents beyond the design basis to provide strategies for severe accident prevention, management, and mitigation.
Plant models are simulated in best-estimate state-of-the-art com-puter codes (e.g., RELAP, TRAC, MARCON-2.08, CONTAIN, MELPROG, SCDAP).
The results will be used to develop better insights into the understanding of complex system transients coupled with multiple system failures, the potential need for automated response, the impact of the human-machine interface on nuclear power safety, the improved understanding of fuel damage and fission product transport phenomena, and the pressure boundary integrity of aging operating plants.
6.1.1 Maior Regulatory Needs and Their Justifications 1.
Analysis of severe accident scenarios for specific types of plant designs, to be used in licensing reviews (1987-1990).
Justification:
A number of postulated high-risk sequences leading to possible severe accidents have been identified by risk assessments of the Accident Sequence Evaluation Program (ASEP) and Reactor Safety Study Methodology Applications Program (RSSMAP).
Detailed analyses of these high-risk sequences are needed to determine appropriate operator actions and any need for special instrumentation.
Specifically recommended opera-tional techniques for managing accident recovery from accident management and human factor research and consequent algorithms to be used by the operator to prevent, diagnose, and respond properly to accidents will be analyzed as a basis for appropriate regulatory actions over the time interval.
6-1 l
l h
2.
Resolution of licensing and safety concerns expressed by NRR in its review of operator guidelines (1990).
Justification: NRR has transmitted to RES for analysis by the SASA program some current safety concerns that arose during the course of licensing reviews and unresolved safety issue (USI) resolution activities.
NRR has requested studies to d9termine the viability of proposed alternative opera-tor actions and the capability of plant systems to restore the plant to a 1
controllable status.
A continuing need exists to support NRR by responding to concerns as they arise. The survivability of equipment from the desig-nation of severe accident environmental conditions is an additional concern to be evaluated by the SASA program.
It is noted that these are safety concerns with respect to public safety but are not licensing concerns for the design basis accident.
3.
Evaluation of information that the operator needs in order to take proper action and evaluation of instrumentation functional requirements to enhance the man-machine information flow when accidents occur, this information to be used in regulatory reviews of existing and proposed instrumentation improvements (1987).
Justification:
From the evaluation of plant-specific response character-istics for a range of accident scenarios, requirements for instrumentation to actuate automated preventive action or to inform the plant operator of the need for manual intorvention will be more fully identified.
Some operational instrumentation systems on operating plants may require re-evaluation as a result of new functional requirements emanating from these studies.
4.
Fission product release and transport assessments for use in equipment qualification, probabilistic risk assessment, and the definition of siting and emergency planning requirements (1989).
Justification:
Fission product release rates are being evaluated for dominant accident scenarios.
This work will assist in assessing the degree of conservatism currently imposed on licensing requirements with respect to fission product transport and source term models.
6.1.2 Research Program Description The strategy is to analyze dominant accident sequences derived from risk assess-ment studies for specific plant designs to evaluate areas of uncertainty and system functional requirements, to assess prevention and mitigation of core melt during severe accidents, to determine appropriate operator actions by performing detailed analyses of high-risk sequences, to evaluate equipment and system survivabilii,y in severe accident environments, and to evaluate the impact of proposed prevention and mitigation features on dominant severe accident sequences.
Test programs such as the Power Burst Facility (PBF), Semiscale, and Full Integral Simulation Test (FIST) will be used to provide definition for issues of concern.
These test programs will also produce data that can be used to evaluate SASA analysis results.
The ASEP will define high-risk sequences for consideration.
NUREG-0956 and NUREG-1150 are to provide a summary of work that is being done by the NRC and its contractors on severe accident source terms.
The ASEP is providing sequence frequency information for five operating
- plants, i.e., Zion, Surry, Sequoyah, Peach Bottom, and Grand Gulf.
The SASA 6-2 n
I program has provided and will continue to provide valuable technical informa-tion in the severe accident risk-reduction program (SARRP) rebaselining and cost-benefit analysis.
Licensing and safety concerns generated by licensing reviews will serve to define SASA issues.
The major research products will be:
1.
Technical bases for improved risk assessments (1987-1990).
2.
Analytical base for licensing rulemaking decisions (hydrogen) (1987).
3.
Basis for detailed licensing decisions (e.g., recent industry proposal and NRC acceptance of manual switchover of reactor core isolation cooling suc-tion in three specific plants) (1987).
4.
Basis for ex-containment source term as distinct from source term release to containment (1987-1990).
6.1.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
The large, dry B&W plant has been analyzed for the potential accumulation of hydrogen in detonable concentrations.
A request for similar analyses of Westinghouse and Combustion Engineering plants is unfunded under planned FY 1986 cuts.
Nor is funding provided for FY 1987.
Planned detonation studies in local areas for equipment survivability loading will not be done.
TRAC-BWR with improved heat transfer will be available in the near future.
Current controversy over whether a BWR core heats up (decay heat only) at low flows was to have been the first major analysis followed by feedback of power (i.e., voiding) effects for high pressure and low pressure (ADS) boiloff (NUREG-1150 studies use constant power).
Elimination of direct support for the fission product transport effort will seriously reduce the capability tc, support a serious effort to do a mechanistic study of the secondary containment fission product retention capability.
NUREG-1150 and industry differ by factors of 3 to 1000.
The secondary contain-ment has recently been upgraded to the status of a major issue in NUREG-1150.
Serious delay will ensue for planned analyses of rec 4uested studies of Grand Gulf and LaSalle pool bypass and of the planned study of LaSalle venting.
December 1985 meetings with NRR resulted in an agreement that a near-term request and a longer-range request for analyses in support of the severe accident imple-mentation program would be forthcoming.
The long range requests are expected as needs are developed by the results of the current request for procurement by NRR for review of research in severe accidents and reduction to aids-for-licensing form.
I Potential 1986 Sandia cuts could terminate:
1.
Hydrogen mixing studies benchnarking (station blackout) comparison with HECTR designed to test transport phenomena by lumped parameter code.
6-3 L
~
2.
Evaluation of MERGE /CORSOR/ TRAP-MELT for correct fission product release and the effect of fission product heating in the upper plenum (report would not be completed until 1987).
3.
Studies of low pressure failure of reactor vessel and resultant coincident hydrogen burn with steam present (1986).
4.
Validation of the MELCOR calculation of Surry station blackout (MELPROG-CONTAIN basis).
5.
Validation of melt progression codes:
(a) with TMI core damage termina-tion point and (b) LOFT tests.
6.2 In-Vessel Core-Melt Progression This element describes research to determine the state of the core during the progression of a severe accident until failure of the reactor pressure vessel The containment loads produced by the ex-vessel core melt are a strong occurs.
function of the core state at vessel failure.
The data base to be produced and the mechanistic codes (MELPROG and SCDAP) to be assessed will provide a tech-nical basis for decisions and actions by NRC concerning accident conditions (and their uncertainties) beyond the current design basis.
These decisions and actions relate to possible changes in regulations and to NRR audits of specific plants.
6.2.1 Major Regulatory Needs and Their Justifications 1.
Determination of the actual hydrogen release from the core and confirma-tion of regulatory action (1987).
Justification:
The rate of hydrogen generation during a degraded-core accident is dependent on the fuel and coolant behavior in the accident sequence.
Evaluation of containment loads and of the adequacy of mitiga-tion and control features requires knowledge of the time and quantity of hydrogen release.
The related amount of unoxidized metals has a strong impact on ex-vessel core-concrete interactions.
2.
Determination of the general behavior of severely damaged fuel in the 1100K to 3000K (1500*F to 5000 F) temperature range for use in imple-mentation of severe accident policy (1987).
Justification:
The recent Severe Accident Policy Statement (see NUREG-1070) requires the industry to search for risk' outliers at operating reactors.
More detailed knowledge is needed on how the core behaves under degraded cooling in order to determine that technical issues and industry analyses resulting from the searches have been resolved and reviewed in a manner to adequately ensure public health and safety.
3.
Determination of the coolability limits and cooling requirements of damaged cores at various stages of degradation, to be used by the NRC staff in reviewing proposed accident recovery and emergency planning procedures (1987).
Justification:
The key question is:
Under what conditions can we be assured the fuel will not melt through the vessel? Data and verified models on the coolability of reactor cores with different degrees of core 6-4 1
damage are needed in order to determine the range'of conditions for which emergency core cooling system (ECCS) reflood can provide accident recovery.
This information is needed for accident management and emergency planning and for risk assessment.
It is also needed for assessing the adequacy of ECC systems and operational plans.
4.
Improvements in vessel failure mode analysis (1988).
Justification:
The mode of vessel failure and the manner with which molten core debris exits the vessel have two major effects on severe accident consequences:
(1) Rapid, high pressure ejection would deliver large amounts of heat directly to the containment, thus potentially causing containment failure, and (2) the debris composition and the timing of its egress from the vessel determine starting conditions for core concrete interactions, which produce aerosols that transport radionuclides.
6.2.2 Research Program Description The strategy is to develop, for a range of accident conditions beyond the design basis, a data base and validated models for assessing the state of a severely damaged core, the hydrogen generation, the coolability of the damaged core by reflooding, and the mode of vessel failure.
The core-melt progression research is the last phase of and will complete the severe fuel damage program outlined in previous versions of the Long-Range Research Plan.
The foundation of the severe fuel damage program is the PBF series (four tests) that were' completed j
in FY 1985.
i
. Work will continue through FY 1987 on postirradiation examination, on analysis, and on reporting of the in-pile tests (with 32 rod bundles) in the PBF at the Idaho National Engineering Laboratory (INEL).
The NRU reactor in Canada can accommodate integral coolant boildown tests with 12-rod full-length bundles.
These tests, conducted by Pacific Northwest Laboratory (PNL), will furnish 4
validation data for the cladding oxidation and hydrogen generation models.
Later tests will be used to study accident recovery by core reflooding.
A series of small scale separate-effect experiments on the governing mechanisms involved in core-melt progression will be conducted in the Annular Core Research Reactor (ACRR) at Sandia and in the laboratory to furnish a data base for model development and assessment. A new set of ACRR experiments, to be conducted out-of pile, are being started to round out the validation of MELPROG models.
Two mechanistic computer codes, SCDAP and MELPROG, will be assessed.
The SCDAP code treats the development of fuel damage in the original core volume, starting with intact rod geometry.
The MELPROG code treats the relocation of liquefied and molten fuel and particulate debris, the attack on the core support plate, core barrel, and reactor vessel, and the conditions of vessel failure.
These two codes furnish a mechanistic basis for evaluating apprcpriate parts of the Source Term Code Package and the advanced risk assessment code, MELCOR.
- Finally, the program consists of benchmark data to be obtained from independent examina-tion of the TMI-2 core and from the CORA facility in West Germany.
The major research products will be:
1.
Report on NRU severe fuel damage full-length tests (1987).
(Also applies to Need 2.)
(
1 6-5 I
2.
a.
Report on analysis of PBF tests (1987).
b.
Final report on independent examination of selected TMI-2 core samples (1987).
c.
MELPROG-BWR code released (1987).
d.
SCDAP-BWR code released (1987).
3.
a.
Report on ACRR debris formation experiments (1987).
b.
Assessment of MELPROG with PBF and ACRR results (1987).
c.
Report on ACRR melt progression separate-effect experiments with irradiated fuel (1989).
4.
a.
Report on MELPROG validation experiments (1988).
b.
Report on limited sensitivity studies using MELCOR/SCDAP/MELPROG (1987).
6.2.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
One NRU test has been deferred until FY 1988 as a result of budget reductions in FY 1986.
This deferral affects the ability to fully confirm core damage predictions for various accident conditions.
Fuel rod materials that have been tested are being held for disposal.
Pending negotiations with the Department of Energy, funds will be required for the disposal activity.
6.3 Hydrogen Generation and Control Research conducted under this program is providing information and analytical models to quantify the loads on containment from hydrogen burning that could exceed the ultimate strength of the building.
The research is providing infor-mation to assess the efficiency of proposed mitigation systems.
This work includes the development of analytical models that will permit better under-standing of hydrogen transport, mixing, and combustion phenomena.
6.3.1 Major Regulatory Needs and Their Justifications 1.
Data from all areas of hydrogen research such as generation, ignition conditions, and mixing to support rulemaking with regard to hydrogen control for LWRs with large, dry containments (1987).t Justification: There is concern for large, dry PWRs that, in the absence of.a hydrogen control system, detonable mixtures might form and detonate.
This program will assess detonability as a function of tempereture and hydrogen-steam concentrations typical of accident conditions in a large, dry PWR.
tA modification of the regulations may result from the research.
6-6 L_-
i i
U 2.
Technical data and information on hydrogen generation and control to help i
formulate the Commission policy on hydrogen regulations not covered by the e'xisting rule (1987).
Justification:
This program will be assessing the threat posed by hydrogen Trom core-melt accidents _more severe than those currently covered by the i
final hydrogen rule.
This information will be used in probabilistic risk l
assessment (PRA) to determine if additional control and mitigation requirements are cost effective.
3.
Specific data on hydrogen combustion phenomena for ice condenser and Mark III containments (1987).
Justification:
In the licensing review of plants with ice condenser and suppression pool types of containments, a number of issues have been l
related to combustion phenomena such as flame acceleration and flame stability.
This research and the research being conducted by Electric Power Research Institute (EPRI) on hydrogen safety will confirm regulatory i
assessments in these areas.
6.3.2 Research Program Description t
As a consequence of an accident, significant quantities of hydrogen can be generated in the reactor vessel from steam-metal reactions and in the contain-ment building from molten-core / concrete interactions.
Burning of this hydrogen leads to pressure loading of the containment.
The hydrogen behavior program is developing improved detonation and diffusion flame models to predict thermal loads and containment pressure histories after hydrogen combustion.
The models include heat transfer by radiation, convection to surfaces, and condensation and evaporation of sprays. Work on understanding the phenomena of flame acceleration and transition from deflagration to detona-tion in containment is being carried out, along with work on hydrogen stratifi-1 cation, mixing, and transport effects.
The experimental portion of the program includes the determination of combustion and detonation limits in air and steam and the effect of the strength and location of the ignition source, geometry, i
degree of venting, and obstacles.
Temperature and pressure profiles as a func-tion of time will be measured.
The effects of hydrogen burning on source term j
attenuation and composition will also be assessed.
A supporting program is studying the prevention and mitigation of hydrogen com-4 bustion by deliberate ignition of lean mixtures of hydrogen with both powered and nonpowered igniters.
Studies include the effects of sprays and condensing steam on igniter performance under accident conditions.
The major research products will be:
1.
Analysis of PWR with large, dry containment for degraded core and severe accidents (1987).
(Also applies to Need 2.)
2.
a.
Assessment of effects of sprays and condensing steam on hydrogen control system and the effects of hydrogen burning on source term (1987).
b.
Hydrogen diffusion model and flame acceleration model (1987).
l t,
6-7 l
'I 3.
See RES Products 1 and 2.a and b.
6.3.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
High-temperature hydrogen-steam mixtures may jet out of breaks or openings from the reactor's primary coolant system into the containment building.
These jets may either ignite spontaneously or a flame may flash back from an ignition source in the containment to the point of release.
The issue is:
Under what conditions will autoignition of jet flames occur for release of a hydrogen-steam or hydrogen jet into an ambient mixture? The detailed diffusion flame modeling effort has been cancelled, thus eliminating the capability to provide a stand-alone model to predict the local environment of safety-related equipment.
In addition, flame acceleration modeling will be discontinued after 1986.
This would have provided a useful predictive tool for assessing potential damage to equipment and structures caused by accelerated flames.
Also unfunded will be the program to assess the effects of condensing steam on the hydrogen control system and the effects of hydrogen burning on the source term.
6.4 Fuel-Structure Interaction Experimental research described in this element will obtain data on the conse-quences of ex-vessel interactions of high-temperature core fuel debris following escape from the vessel in severe accidents.
The types of interactions of concern are thermal and chemical interactions between core fuel and (1) the reactor cavity concrete basemat, (2) water present at the time of fuel debris escape or subsequently introduced to the cavity, and (3) the containment atmosphere.
6.4.1 Major Regulatory Needs and Their Justification Experimental research and analytical studies are needed to assess characteristic interaction responses for:
1.
Heat generation and release for analysis of containment performance during severe accidents (1987).
2.
Noncondensible gas and aerosol release for analysis of containment perfor-mance during severe accidents (1987).
3.
Rapid steam generation with potential for containment failure (1987).
4.
Rapid pressurization of the containment in high pressure melt ejection with potential for early containment failure (1987).
Justification of Above Needs:
Current assessments of the margins of safety to containment failure under core-melt sequences have overlapping uncertainty bounds between load and response estimates.
Thus, the primary need for this experimental research is to develop data upon which better quantitative assessments of the challenge to the containment structure can l
be made for postulated severe accidents involving release of core debris to the reactor containment.
This applies to the four needs listed above.
6-8 l.
The sources for the containment challenge are not adequately known, and l
more research is needed to evaluate (1, 2, and 4) effects of the inter-action between hot fuel debris and concrete basemat materials and between core debris and the containment atmosphere; (3) effects involving rapid steam generation; and (1-4) the quantification of parameters used to provide a basis for establishing and verifying analytical models used in l
severe accident assessment and accident management planning.
6.4.2 Research Program Description The plan of research is to conduct small-and large-scale scoping and phenomeno-logical tests of core melt / concrete, core-melt / atmosphere, and core-melt / coolant interactions in crder to quantify analytical models for noncondensible gas and aerosol sources and heating of the containment atmosphere.
In addition, the effects of introducing coolant to the fuel-melt mass will be evaluated.
Hot solid interaction tests will be performed to assess long-term cooling problems of solidified melts.
For steam explosions, the research program will address the major uncertainties of mixing, triggering at high pressures, energy conversion ratio, and fraction of melt participating.
Data analysis and model development will comprise about one-third of the program.
In cooperation with the United Kingdom Atomic Energy Authority (UKAEA) at Winfrith, program guidance will be provided annually by a Research Review Group (NRR, RES, and advisors).
One program principle will be that the theoretical efforts provide guidance for the experimental efforts and conversely.
The experimental program will be restricted to less than 100 kg thermite melt per test for the immediate future.
The major research products will be:
1.
Verification of core melt / concrete interaction models in CORCON to be obtained from large-scale test facilities (1987).
2.
a.
CORCON and CONTAIN code verification (1987).
b.
Experimental data for modeling the interaction of hot solidified melts with concrete and long-term cooling characteristics of solidified core melts (1987).
c.
Predictive mechanistic models for mixing and explosive interaction of fuel melts with water in pouring and stratified contact modes (1987).
(Also applies to Need 3.)
3.
Experimental tests from high pressure ejection tests to quantify:
Dispersion of core debris out of reactor cavity (1987).
a.
b.
Debris particle size distribution and aerosol generation (1987).
c.
Rapid pressurization of containment due to thermal and chemical interactions of dispersed core debris with containment atmosphere (1987).
6-9
4.
For steam explosions, predictive calculations using newly developed models in existing computer codes to scale experimental results to reactor scenarios (1987).
6.4.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
The research needs remaining for steam explosions can be dividad into three areas:
1.
Quantification of the liquid water starvation in the interior region of multiple melt jets for a pouring contact mode potentially resulting in a low-energy yield.
2.
Quantification of the depth of self-mixing of the corium and water for a stratified contact mode potentially resulting in insufficient mixing and a low-energy yield.
3.
Quantification of the total energy yield from multiple explosions in the lower plenum potentially resulting in a low-energy yield.
Preliminary calculations show that the volume of the mixing region interior to multiple melt jets would contain almost all steam and little liquid water.
The implication is that liquid corium/ liquid water contact would occur almost exclusively along the outer perimeter faces of the multiple melt jets, a rela-tively small surface area.
Given this water-starved central region, the resultant energy conversion ratio is expected to be low and the overall energy yield to be less than 2000 MJ.
A small number of experiments could quantify the effect of this phenomenon and provide sufficient information for predictive models to confirm the low-energy yield and the resultant elimination of alpha failure consideration for the pouring contact mode.
Research results by Kottowski show that the depth of self-mixing of the corium and water in a stratified contact mode can be 15 centimeters.
Preliminary calculations show, assuming a conversion ratio of 10 percent, that in the lower plenum of the reactor the required depth of self-mixing is 27 centimeters for an energy yield of 2000 MJ.
This mixing depth is not impossible but is not expected since the self-mixing and the explosion have to simultaneously occur within a small fraction of a second.
A small number of experiments could quantify the effect of this phenomenon and provide sufficient information for predictive models to confirm the low-energy yield and the resultant elimination of alpha failure consideration for the stratified contact mode.
Los Alamos National Laboratory calculations with the SIMMER code show a force of about 1.5 GN from simulated multiple incoherent steam explosions. The UKAEA Winfrith experiments show that multiple explosions occur predominately, given their underwater release conditions.
If the expected results are obtained for the first two issues, the issue of multiple explosions is expected to be solved by calculating the energy yield obtained for bounding reactor conditions.
6.5 Containment Analysis This element provides analytical tools for the assessment of the challenge to the reactor containment system from postulated severe accidents.
The types of 6-10
challenges produced by postulated accidents may consist of overpressure from steam generation and noncondensible gases, fission products, and aerosol releases; hydrogen burns; fuel-structure interactions in the reactor vessel cavity; and direct heating of the containment atmosphere by the core debris discharged from the reactor vessel under high pressure.
The analytical tools developed by this program will be used to support the SASA program and to assist in regulatory review of containments.
6.5.1 Major Regulatory Need and Its Justification The CONTAIN code, a comprehensive, integrated systems code, has been developed to provide an analytical tool for assessing the loading imposed on containment systems by severe accident conditions.
The code is structured so that new models, data, and special purpose computer codes can be incorporated into CONTAIN as they are developed.
It serves as a foundation for integrating the products of the containment loads research program into an interactive system code.
Results of many experimental programs are being used for code validation and for improve-ment of the models used in the code.
Development of the CONTAIN cc,de will continue in order to incorporate future experimental and analytical results.
In order to identify areas where large uncertainties exist in the analysis of containment loading, a quantitative uncertainty analysis is needed.
Justification: The containment analysis task is the key to integrating the research on severe accident phenomenology into a tool that can be used for regulatory audits of containments and that can be used to provide an improved basis for risk analysis.
The CONTAIN code is now being used by the Sandia, Idaho, and Oak Ridge laboratories for SASA.
Decisions on future regulatory actions dealing with severe accidents will require a state-of-the-art capabil-ity for analyring containment performance.
6.5.2 Research Program Description This research undertakes to develop a computer code capable of simultaneously assessing the many-faceted challenges to containment during severe accidents.
The effort to incorporate phenomenological models for the many simultaneous processes occurring within the containment system into a single code structure continues.
Some of the safety analysis computer codes that have been integrated into CONTAIN are (1) MAEROS for aerosol calculations; (2) CORCON/VANESA for core /
concrete interactions, including aerosol production, fission product chemistry and release rates, gas production, penetration, and atmospheric heating; (3)
HECTR for hydrogen behavior; (4) ICEDF for aerosol behavior in ice condenser containments; and (5) MEDICI for debris behavior and SPARC for BWR suppression pools.
The major research products will be:
1.
a.
CONTAIN code (1) State-of-the art containment systems code, CONTAIN, incorporating best mechanistic containment analysis tools (1987).
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(
(2) Improvement and verification of aerosol models based on Karlsruhe (Germany) DEMONA, EPRI-LACE, and ORNL Nuclear Safety Pilot Plant test data (1987).
(3) Integration interface between CONTAIN code and other codes such as TRAP-MELT for fission product, energy, and material source data (1987).
(4) CONTAIN code validation (1987-1990).
(5) Application of CONTAIN code to specific licensing cases (1987-1991).
b.
Quantitative Uncertainty Evaluation of Containment Loading Analysis (QUECLA)
(1) Methodology for identifying and ranking important sources of uncertainty in containment loading analysis (1987).
(2) Evaluation of uncertainties for several plants and accident sequences to delineate key phenomenological contributors to uncertainty and to provide quantitative ranking of their importance (1988).
6.5.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
Several areas of containment systems research need to be incorporated into the CONTAIN code when experimental work is completed.
If funds are available, this will be done in FY 1987 through FY 1989.
These areas are:
1.
Direct heating of containment atmosphere by high pressure discharge of corium following vessel meltthrough.
2.
Effects of moisture and steam on aerosol behavior.
3.
Three-dimensional effects on hydrogen stratification and on intercell transport of aerosols.
6.6 Fission Product Release, Behavior, and Transport The fission product and aerosol release, behavior, and transport research program is directed toward assessing computer models and obtaining experimental data under realistic accident conditions to determine the potential radio-logical source term released from LWR plants during severe accidents.
The research should enable an improved understanding (e.g., reduced uncertainty) of the source term.
Source term calculations are needed to support implementation of the severe accident policy statement.
This research includes studies on radionuclide release from the fuel, on transport and attenuation of radio-nuclides and aerosols within the reactor coolant system, and on attenuation (depletion) within the containment vessel.
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6.6.1 Major Regulatory Needs and Their Justifications 1.
Determination of amounts, chemical forms, and aerosol characteristics of fission products released in-vessel and ex-vessel and removed in the reactor coolant system and containment under severe accident environments, for use in confirming regulatory decisions concerning conditions beyond the design basis (1987).
Justification:
Risk assessments, risk reduction efforts, and implementa-tion of the severe accident policy statement (see NUREG-1070) depend on an improved understanding of the release of radioactive materials to and from the containment.
Major uncertainties still exist with regard to iodine chemistry in a reactor environment, tellurium behavior, potential in-vessel and ex-vessel releases of plutonium, and the various retention mechanisms affecting all of these.
2.
Improvements in PRA consequence calculational methods (1988).
Justification:
PRAs are required for new plant construction applications.
The present state of risk assessment techniques suffers from limitations both in the methodology and in the incompleteness of the phenomenological data base.
A better understanding of the phenomenology of accidents involving fuel damage and the corresponding fission product release will permit realistic treatment and should much improve the usefulness of risk assessment.
6.6.2 Research Program Description The strategy of the program is to develop an experimental data base, scientific information, and computer codes for predicting (with reduced uncertainty) release, chemical form and reaction, deposition, and transport behavior of radionuclides under severe LWR accident conditions. This will allow the technical bases for licensing, inspection, and regulatory practices to be substantiated.
Computer models are being validated to assess fission product and aerosol release from the fuel during the in-vessel heatup and melting phase and (as discussed in Section 6.4) during ex-vessel interactions of fuel debris with reactor cavity materials (e.g., concrete).
Models are being validated and improved to assess the transport, reaction, and deposition of radionuclides within the reactor coolant system components and piping and within the main containment compartment (s).
Models for quantifying the effectiveness of engineered safety features (ESFs) in mitigating fission products under severe accident (fission product and aerosol) loadings and environmental conditions are discussed in Section 6.8.
Some of these models and codes are used as part of the Source Term Code Package and some as separate detailed computer codes.
The latter are used in best-estimate source term analyses for benchmarking the MELCOR code.
Laboratory-scale separate-effect experiments are under way to provide data for model validation in a number af areas, including fission product release from fuel, fission product chemistry, fission product interactions with prototypic surface materials, and fission product aerosol behavior in the reactor coolant 6-13
system and containment.
Data on the behavior of radionuclides in the lanthanide and actinide series are of particular importance.
Completed experiments (e.g., PBF) are being analyzed and additional experiments are being conducted to assess the validity of the computer models.
These experiments include the large-scale in pile fission product release experiment to be carried r.ut in the NRU.
Using ^.he models developed within this element, periodic analyses will be con-ducted to refine best-estimate, release-from plant, radionuclide source terms for severe LWR accident sequences.
A major source term reassessment was completed in 1986 (see NUREG-0956).
The major research products will be:
1.
a.
Data report for validating and improving in-vessel fission product release codes (FASTGRASS and VICTORIA.. from out-of pile tests up to 2700K (1987).
b.
Data report for comparing out-of pile experimental results on in-vessel release of fission products with in pile results (1987).
c.
Summary report on fission product releases measured in PBF and NRU tests (1987).
d.
Large-scale aerosol transport test results from Nuclear Safety Pilot Plant to be used to assess TRAP-MELT code and CONTAIN code (1987).
2, Severe accident source term uncertainty report (1987).
6.6.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
Models and codes for high burnup (>60,000 MWD /MTM) fuel or for recycle (M0x) fuel need to be validated.
Although the amount of fission product release has been determined, posttest analyses of residual fuel specimens from in pile and out-of pile experiments are needed to determine the mechanisms of that release.
Completion of the separate-effect experiments in the ACRR (ST series) is not being funded in FY 1987.
6.7 Containment Failure Mode This element treats three possible failure modes:
valve failure, materials failure in electrical penetratinns due to high temperature, and mechanical failure of the containment due to either excessive local deformation at major penetrations or structural failure.
Both assessments of the risk posed by loads outside the design basis, such as hydrogen burns or basemat melting, and estimates of the effectiveness of proposed mitigative steps require an ability to predict the way in which a containment could fail.
However, this element does not address the failure mode arising from the failure to isolate the con-tainment because of improper valve positioning.
Both the utilities and the NRC address this part of the problem through quality assurance practices, inspection and enforcement, and other administrative and management techniques.
Scenarios that bypass containment via penetrations are also treated in Section 6.1.
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6.7.1 Major Regulatory Needs and Their Justifications 1.
The capability to predict, with a high degree of confidence, the pressure and temperature environment that can be sustained by any of the great variety of containment structure designs before the rate of leakage becomes unacceptably high (1988).
Justification:
State-of-the-art methods cannot reliably predict whether leakage will begin around penetrations or in the membrane region of the shell.
If leakage at penetrations is critical, the effects of aging on gasket performance will be of significance.
The technical problems involve developing an ability to predict deformations for the wide array of containment types and relating deformations of containment structures to leak behavior.
In addition to failures of the shell structure or penetrations, possible leakage paths exist through isolation valves and electrical penetrations.
The staff must have the capability to evaluate the leak rate estimates made of the capabilities of a wide range of containment designs.
2.
The development of simplified computational models, suitable for use in risk analyses, that adequately represent the variability of containment L
performance under severe loadings (1987).
Justification:
The implementation of a safety goal would require, as part of the PRA calculations, computational models describing the performance of containments.
In particular, the implementation of a containment per-formance criterion, i.e., conditional probability that a containment will function given an accident, would require an ability to relate variability in leakage behavior to variability in structural parameters and accident t
conditions.
3.
An ability to asess the extent to which containment performance (leak-tightness) may be degraded in accidents initiated by extreme external events such as a major earthquake (1990).
Justification:
The first. generation of PRAs for nuclear power plants indicates that severe environmental events are likely initiators of severe accidents.
Effectively, the current practice is to assume that contain-ment performance is the same, whatever the sequence of events leading up to a severe accident.
More realistic models of containment performance must consider external events such as an earthquake.
These may result in the failure of penetration seal or gasket materials or in distortion within the penetration areas large enough to destroy the leak-tightness of the containment.
In turn, the results of a severe accident resulting from the external event may not be sufficiently contained.
6.7.2 Research Program Description The research effort will focus on five areas:
1.
Model tests of containment structures aimed at verifying computational methods for predicting deformations and failure.
2.
Experiments to understand the behavior of seals and gaskets when subjected to severe accident conditions.
3.
Experiments on models of penetrations to relate leakage behavior to local deformations and pressure-temperature environments.
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4.
Experiments on leakage behavior of electrical penetrations in high-temperature environments.
5.
Experiments on the performance of containment valves when subjected to severe accident loadings.
There is, and will continue to be, significant interaction with other NRC-sponsored programs related to the severe accident research program.
Particu-larly close coordination will be maintained with the programs on hydrogen gen-eration and control, fuel-structure interaction, and containment analysis (see Sections 6.3, 6.4, and 6.5).
In addition, there will be interactions with the risk code development program.
There will also be interaction with other national and international programs.
Contributions to this program from EPRI-sponsored work are anticipated in the way of analytical predictions of capacity to be compared against test results and in the localized behavior of concrete containments under severe accident loadings.
Two foreign programs have been identified as sources of information.
One is c
the effort on prestressed concrete containments being conducted in France. The other is the planned testing, on a shake table in Japan, of containment models to simulate seismic response.
Experiments involving tests of steel containment models under static pressure were completed in 1985.
Reinforced concrete models will be tested in 1986 and 1987.
The tests to understand the behavior of seals and gaskets under simulated acci-dent pressure-temperature profiles were being performed in the 1984-1986 period.
The shell penetration interaction effects have been studied in steel model tests and will be studied in concrete model tests.
Other large-scale testing of penetrations (containment airlocks, equipment hatch) were planned in 1985 to be performed in 1986-1987.
Planning for simulated seismic testing of containment models began in 1986.
The actual testing depends, in great measure, on the extent of cooperation developed with the Japanese research program on seismic testing. The three options currently under consideration are cooperative testing using the large Japanese shake table facility at the Nuclear Power Engineering Test Center in Japan; simulation of earthquake ground motion by phased explosive arrays; and quasi-dynamic loading using hydraulic actuators. The first seismic tests are anticipated in 1988.
Valve performance tests and tests on electrical penetrations were performed in 1984 through 1986.
The major research products will be:
1.
a.
Results of tests on models of major penetrations (1987).
b.
Comparison of predicted capacities for prestressed and reinforced concrete containments with experiments under static pressure (1987).
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2.
Comparison of predictions of capacity for reinforced and prestressed concrete containments under dynamic pressure loads with experimental results (1987).
3.
Results of initial tests of containment models under simulated seismic loading (1988).
6.8 Fission Product Control The fission product control program was developed to evaluate the effectiveness of ESF systems under severe conditions as a part of broad research needs to support the reassessment of the regulatory assumptions of severe accidents and to develop a technical basis for decisions concerning the response of ESF systems to the source term reassessment and to severe accidents.
6.8.1 Major Regulatory Need and Its Justification Validation of the effectiveness of ESF systems under realistic estimate of revised fission product source terms to provide input for (1) severe accident policy and regulation questions and (2)isource term reassessment for basic LWRs (1487).
Justification:
Evaluation of the impacts of revised source terms on the do in and effectiveness of ESF systems for a spectrum of accident conditions is needed as part of the information base for formulating policies and strategies to mitigate the postulated fission product loadings in a severe accident.
A concern arising from past regulatory emphasis on emitted radioactive iodine is that this practice may have resulted in a misplaced emphasis on ESF-system design.
A review of mitigative ESF systems used in cu rent LWR designs shows that the combination of ESF systems used in contemporary power reactors results in effective mitigation of all currently postulated accident sequences within the design basis accident (DBA) envelope.
For the DBA, conservatisms exist in the form of simplifying assumptions and underestimates of some of the ESF-system effectiveness.
Most ESF systems are likely to be functional for postulated accidents substantially more severe than the DBA.
There is, however, substan-tial variation in tne effectiveness of fission product removal of various ESF systems under conditions exceeding their design basis.
The results of this research are expected to produce significant new informa-tion that will permit evaluating ESF-system design and effectiveness for the full spectrum of accidents and are therefore expected to contribute to future regulatory decisions.
6.8.2 Research Program Description The strategy for the research is to obtain and develop technical information that will assist in providing best estimates of the spectrum of chemical and physical properties of the severe environments expected to be imposed on the ESF systems and to evaluate and predict ESF-system performance under such conditions.
s The research is closely coordinated with the other NRC severe accident programs as well as with those conducted by other foreign countries'and the United States nuclear industry (EPRI, General Electric, Westinghouse).
The existing and expected research results are and will be extensively used to achieve the objectives of this program.
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- 1
Research will concentrate on the prediction of the extent of the removal effectiveness and the depletion of aerosols and other fission products by ESF systems such as containment sprays, suppression pools, ice beds, and filter systems; on the quantification of the effectiveness of ESF and other mitigation features in reducing the potential fission product escape from containment; on the evaluation of hydrogen burning on the performance and the effectiveness of ESF and aerosol concentrations under such conditions; on an evaluation of the existing design features under expected aerosol loadings; and on the develop-ment of simulated conditions and design and operational features of ESF for a generic evaluation for standardized nuclear facilities.
Codes will be developed and verified for ESF-system reliability, accelerated aging, and evaluation of safety and technical benefits as well as cost benefit for alternatives to some of the existing ESF systems.
The major research products will be:
1.
a.
Verification based on experimental data to be provided by EPRI (1987).
b.
Code (ICEDF) verification for evaluation of PWR ice-condenser effec-tiveness (1988).
c.
Code modification and verification for performance and effectiveness of PWR/BWR containment sprays under severe accident conditions (1987).
d.
Code development and verification for evaluation of generic design of ESF systems for standardized nuclear facilities (1987-1989).
6.9 Severe Accident Risk and Risk-Reduction Analysis This program provides for an integrated assessment of the risks presented by current generation LWRs and related analyses of the costs and effectiveness of proposed risk-reduction measures.
The principal near-term objective of this program is to provide an integrated reactor risk assessment report that would assemble and interpret the most recent information on accident irequencies and on the behavior of fission products and reactor containment systems during severe reactor accidents. This report will integrate the results of the experi-mental program detailed in Sections 6.1 through 6.8 and reflect the most recent advances in data analysis and risk assessment M thods and code development as set forth in Sections 4.1 and 4.2.
The specific details of this program element, including specific research products, are described in Section 4.2 of Chapter 4 of this plan.
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1
7.
RADIATION PROTECTION AND HEALTH EFFECTS The goal of the NRC's radiation protection program is to ensure that workers and members of the general public are adequately protected from the adverse consequences of exposure to ionizing radiation from licensed activities.
RES activities supporting this goal include developing radiation protection standards and guidelines for the implementation of these standards and con-ducting research studies to provide the information necessary for the standards development process.
This includes resolving uncertainties or inconsistencies in current information and filling in existing gaps in the knowledge required to ensure an adequate foundation for NRC standards.
7.1 Metabolism and Internal Dosimetry 7.1.1 Major Regulatory Need and Its Justification 1.
Metabolism model for thorium (1989).
Justification:
Workers at thorium refineries are exposed to thorium com-pounds via inhalation.
In order to set limits for concentrations in air, it is necessary to calculate doses to body organs (principally the lung) following radionuclide incorporation.
Current information on the metabo-lism of incorporated thorium is very limited so that it is necessary to study individuals who have been exposed to thorium in order to determine the metabolic parameters.
7.1.2 Research Program Description Mortality data through March 1984 on causes of death for male thorium workers will be analyzed by appropriate statistical techniques.
Pulmonary function data will be analyzed to determine the effect of thorium on the lung.
Radiochemical analyses of remaining autopsy materials on hand and any new specimens will be analyzed for thorium and daughter concentrations.
The major research products will be:
1.
Standard mortality ratios for thorium workers (1987).
2.
Metabolic parameters for thorium (1988).
3.
Metabolic model for the metabolism of thorium (1989).
- 7. 2 Health Effects Standards Development 7.2.1 Major Regulatory Need and Its Justification 1.
Regulations for radiation protection (1987).
Justification:
Standards to protect radiation workers and the general public from exposure to radiation are key elements of NRC's regulatory program.
It has been over 20 years since NRC's 10 CFR Part 20 was issued.
7-1
New data and a new system of dose limitation have been developed and are being used in other countries.
It is necessary for the NRC to adopt ieg-ulations that reflect current national (especially the Federal Radiation Guidance developed by the Environmental Protection Agency) and interna-tional (ICRP-26) recommendations.
7.2.2 Research Procram Description The first version of the revised 10 CFR Part 20 was published in January 1986.
More than 300 comment letters were received.
The principal activity in FY 1987 will be the modification of the proposed rule to reflect public comments.
The major research product will be:
1.
10 CFR Part 20 final rule (1989).
7.3 Occupational Radiation Protection 7.3.1 Major Regulatory Needs and Their Justifications 1.
Reports of occupational radiation doses received at certain NRC facilities (annual).
Justification:
NRC must look for trends in radiation exposure in order to decide whether further regulations are necessary.
Also the NRC maintains individual termination reports to facilitate the acquisition of occupation-al dose histories.
2.
Evaluation of solidification processes and containers used for decontam-ination solutions (1989).
Justification: Worker exposure due to radioactive waste handling is of concern to the NRC.
Evaluation of dose-reduction technology can contrib-ute to reducing exposures to waste handlers.
7.3.2 Research Program Description 1.
Annual statistical summary data submitted by licensees will be collected and entered into a corrputerized system called REIRS (radiation exposure information reporting system).
2.
If solidification of decontamination solutions leaves free water or if there is an interface problem between the wastes and their containers, radiation exposures can be increased.
Laboratory tests on solidification processes and containers used for decontamination solutions will be performed.
The major research products will be:
1.
Report on occupational exposure (Annual).
2.
Report on waste solidification processes and containers (1989).
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8.
WASTE MANAGEMENT Regulation of radioactive waste management requires a technical capability to assess compliance of a waste management system with the regulatory requirements for operational safety, occupational radiological protection, and long-term waste isolation.
8.1 High-Level Waste High-level-waste (HLW) management includes the regulation of operational safety, occupational radiological protection, and the long-term isolation of HLW.
The Department of Energy (DOE) has the responsibility to design, construct, and operate an HLW repository and to demonstrate that it complies with the standards and regulations established or to be established by the Environmental Protec-tion Agency (EPA) (40 CFR Part 191) and the NRC (10 CFR Part 60) and in accord-ance with the Nuclear Waste Policy Act of 1982.
Regulation of geologic disposal of HLW requires that NRC perform an independent assessment of DOE compliance sufficient to provide reasonable assurance of safety.
Such an assessment must be based on a thorough understanding of the relevant phenomena and processes that affect the performance of a geologic repository both during and after waste emplacement operations.
Effective regulation also requires providing timely guidance to the DOE, especially in consultations during the time prior to submittal of a license application.
As the results of NRC HLW research become available, they are provided to the licensing staff to aid in providing such guidance to the DOE.
8.1.1 Major Regulatory Needs and Their Justifications 1.
Evaluation of Releases to Accessible Environment--capability to evaluate DOE's safety analysis reports (SARs) to assess compliance with 10 CFR Part 60 and 40 CFR Part 191 (release to the accessible environment), to be used in reviewing DOE's license application (1990).
Justification:
Research into the physical phenomena relevant to reposi-tory performance, including investigation of the scientific and technical bases for the DOE performance assessment, is needed to enable NRC to evaluate DOE's demonstration of compliance with 10 CFR Part 60.
Included is research into methods to allow the licensing process to deal with the uncertainties associated with predicting future changes to the site that might affect waste isolation.
Part 60 requires that the SARs contain descriptions and analyses of the repository site's geologic barrier to radionuclide migration.
The hydrology, geology, and geochemistry of the site will have to be analyzed.
Research is needed so that NRC's licensing staff will understand how ground water transports radionuclides through the geologic barrier to the accessible environment and how geochemical effects can retard radionuclide migration.
2.
Evaluation of Containment Requirement--capability to assess the DOE demon-stration that HLW packages will comply with long-term radionuclide contain-ment requirements defined in 10 CFR Part 60, as part of the assessment of DOE's license application (1990).
8-1
Justification: The potential hazard posed by HLW will last thousands of years.
It is NRC's policy that containment of HLW must be substantially complete during the period when radiation and thermal conditions in the underground facility are dominated by fission product decay and that any release of radionuclides from the engineered barrier system will be a gradual process that results in small fractional releases to the geologic setting over long periods of time.
(See performance objectives in 10 CFR Part 60.)
Research is needed to understand mechanisms of waste package degradation and failure and to identify the uncertainties associated in predicting both the behavior of waste form and package material in repository environments and the nature of those environments.
The results of this research will facilitate an independent NRC assessment of the validity of the methods and tests used by the DOE to predict long-term waste package performance.
3.
Evaluation of Release Rate Requirement on Engineered Facility--capability to assess DOE's demonstration that the engineered facility and waste package will comply with the release rate criterion of 10 CFR Part 60 as part of the assessment of DOE's license application (1990).
Justification:
10 CFR Part 60 requires that following loss of containment the release rate of radionuclides from the underground facility must not exceed one part in 105 per year of the inventory present 1,000 years af ter permanent closure.
To perform an independent review of the DOE's predic-tion of repository performance, the NRC needs to know and understand the phenomena that control the rate of radionuclide release from the facility.
Among the phenomena of concern are the impact of thermal perturbations on the geochemical environment, the hydrological flow conditions proximate to the emplaced wastes, and the functioning of packing and backfill materials to control influx and chemistry of ground water and sorption of radionuclides.
Research is needed to understand the chemical process by which radionu-clides enter the ground-water system of a repository.
Since there will be no opportunity to observe actual dissolution leaching of radionuclides from a waste fc m over very long periods, the results of short-term tests will have to be extrapolated.
Further, the actual waste form will have undergone an aging process before the leaching process is expected to begin.
As with waste package failure mechanisms, understanding of experi-mental and testing methods is needed to allow confident review of scaling to the long periods required for acceptable repository performance.
Of particular relevance to extrapolating laboratory experiments to the long periods of time required for repository performance is research on natural systems and materials exposed to conditions analogous to specific aspects of the repository environment for very long times--times on the order of those required for isolation of HLW.
Study of these " natural analogues" can increase our confidence that NRC licensing tools and 00E analytical and field studies will be adequate representations of expected repository performance.
i 8-2
L 8.1.2 Research Program Description The strategy of the HLW research program is twofold:
(1) to identify and develop an understanding of basic phenomena and processes that determine the performance of an HLW geologic repository and (2) to assess the uncertainties associated with predicting repository performance using models based on that understanding.
Hence, such identification and understanding form the technical bases for assessing DuE license submittals for construction and operation of HLW reposi-tories so that the NRC can determine whether there is confidence that long-term performance objectives of 10 CFR Part 60 and 40 CFR Part 191 will be met.
Coordinated program efforts of laboratory and field experimentation and theoret-ical studies will provide both the identification and understanding of the t
processes and conditions that control the long-term performance of the system and the assessment of uncertainties inherent in modeling repository performance.
Relevant research sponsored by the DOE, Electric Power Research Institute, Department of Interior, Nuclear Energy Agency of the Organisation for Economic Co operation and Development (OECD), and foreign governments is being factored j
into the planning of the NRC waste management research program in order to avoid unnecessary duplication and to maximize research effectiveness.
An essential complement to understanding the relevant phenomena is the identiff-cation and assessment of uncertainties pertaining to both the performance and i
assessment of performance of a geologic repository.
Basically, the issue is the degree of confidence attainable in predicting long-term repository perfor-mance that is ultimately based on extrapolation (through models) from short-term, small-scale laboratory and field tests and observations.
Included are 1
- onsiderations of waste package and engineered system performance (e.g., over-l pack corrosion, failure of weldments, waste form leaching); geochemical inter-actions with respect to radionuclide migration and ground-water transport of radionuclides (e.g., chemical speciation, absorption, diffusion); effects on geological stability and isolation from excavation and impacts of emplaced wastes (e.g., thermal propagation of fractures, resaturation, performance of shaft seals); response of the engineered barrier system to changes in the i
repository environment (e.g., climatic changes, resaturation, wet-dry cycling, coupled interactions of thermal-mechanical-chemical-hydraulic systems); and models used to predict overall system performance.
The HLW research program l
also addresses monitoring methods and instrumentation reliability.
Waste management research is an iterative process.
Phenomenological research leads to conceptual models that direct the focus of further phenomenological research.
Moreover, fundamental work in understanding important HLW phenomena is of ten factored into the regulatory program through predictive models. An essential determinant of the usefulness of a model in the licensing process is 4
its reliability as established by field validation work that both confirms and
{
clearly defines limits of utility of the model.
In addition to establishing the models of individual components and phenomena, a systematic and comprehensive integration of the understandings supporting the models must be carried out to ensure a uniform and complete approach to assess-ing repository performance.
This " systems integration" is key to ensuring that RES resources are allocated according to level of uncertainty and importance to i
repository performance--there is no point to intense study of trivial problems.
{
1 8-3
\\
Nonetheless, because of budget constraints, systems integration work has been deferred indefinitely.
Finally, off-normal conditions that stress repository systems beyond their design need to be studied to probe the safety margin in the natural and engineered repository systems and ensure that the repository will perform acceptably even under excursions from expected repository condi-tions without application of undue conservatisms.
The major research products will be:
1.
Evaluation of Releases to Accessible Environment a.
Identification and description of important parameters and functional relationships for hydrological and geochemical models for assessing radionuclide transport processes (1988).
b.
Report on field validation studies of radionuclide transport model (1990).
c.
Parameters and processes important to evaluation of effectiveness of borchole plugging and sealing and shaft sealing techniques for salt (1988).
d.
Parameters and processes important to evaluation of effects of cyclical re-wetting in tuff on borehole and shaft seals (1989).
e.
Assessment of effects of likely off-normal conditions that might result in degraded repository performance (1990).
2.
Evaluation of Containment Requirement a.
Identification and description of the relationship among parameters important in container manufacturing and expected long-term container performance (1987).
b.
Identification and description of methods for predicting long-term performance of waste packages in tuff (including waste form, container, and overpack) (1989).
c.
Identification and description of waste package performance in the environment of a salt repository (1990).
3.
Evaluation of Release Rate Requirement on Engineered Facility a.
Parameters and physical relationships important to methods for evaluat-ing long-term performance of backfill systems proposed by DOE (1988).
)
b.
Applicability of hydrothermal and geothermal data and predictive tech-niques to waste isolation performance assessments (1987).
c.
Identification and description of important parameters and functional relationships for hydrological and geochemical models for assessing radionuclide transport processes (1989).
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d.
Report on field validation studies of radionuclide transport model (1990).
8.1. 3 Deferred /Unfunded Research Needs (as of February 1, 1986)
The waste management research effort in HLW is attempting to maintain progress in three distinct geologic media (basalt, tuff, and salt) and three subject areas:
Evaluation of Releases to Accessible Environment, Evaluation of Contain-ment Requirement, and Evaluation of Release Rate Requirement on Engineered Facility.
However, at present funding levels, there are insufficient resources to adequately address all the technical issues sufficiently well to confidently support projected licensing schedules.
Specifically, systems integration work has been removed from the list of projected accomplisnments, work in basalt is being prematurely terminated, and critical work on field validation of flow and transport models is being initiated at a reduced level, anticipating higher budget levels in future years.
The termination of research on basalt could result in licensing delays if a site is selected in basalt because some issues will lack sufficient understanding for resolution until well into the licensing review.
Other topics that have been deferred include waste package performance in salt, borehole and shaft sealing in tuff, cyclical re-wetting effects on radionuclide transport, and repository response to off-normal conditions.
In each case, the availability of information in these areas is important to the NRC licensing review.
Inadequate funding levels will result in greater uncer-tainty on critical licensing issues and could lead to licensing delays.
8.2 Low-Level Waste Low-level-waste (LLW) management includes occupational radiation protection, protection of the general population from releases of radioactivity, and environ-mental protection.
Land disposal of LLW must be dealt with in compliance with the NRC's 10 CFR Part 61.
Congress mandated in the Low-Level Radioactive Policy Act (Public Law 96-573) that each State have the capability of disposing of LLW generated within that State by 1986.
Subsequent legislation in the form of the Low-Level Radioactive Waste Policy Amendments Act of 1985 extended the period within which States and compacts must establish low-level radioactive waste dis-posal facilities although schedules and penalties for not meeting those schedules were included.
In addition, the legislation charged the NRC with the task of determining which alternative technologies (alternative to conventional shallow-land burial (SLB)) for LLW disposal are appropriate and providing information to the States and compacts needed for licensing such alternative technologies.
The Amendments Act also charged the NRC with licensing disposal of wastes with higher concentrations than acceptable for Class C (above Class C wastes), which will be the responsibility of the DOE.
The NRC will license disposal in non-Agreement States.
Disposal of low-level wastes licensed by Agreement States must be carried out in a manner compatible with NRC safety requirements.
In order to determine compliance with 10 CFR Part 61 (and in particular to evaluate the long-term performance of specific disposal facilities, including alternative technologies and disposal of above Class C wastes, and to provide the guidance required by the Amendments Act),
8-5
the NRC must know and understand the phenomena and processes that affect the stability of disposal sites, the isolation and movement of radionuclides, and the associated uncertainties that enter into any demonstration that release criteria are met.
8.2.1 Major Regulatory Needs and Their Justifications 1.
Methods and procedures for determining the long-term stability of packaged LLW and analyses of the long-term performance of disposed wastes, to be used in evaluating waste form and long-term performance (1987).
Justification:
Section 61.56 of 10 CFR Part 61 specifies a series of mini-mum requirements to ensure LLW stability. The regulation is written in general terms and does not provide detailed prescriptive requirements.
Research is needed to develop guidance and acceptance criteria for how the stability requirements can be met over the design life of the waste packages.
Also needed is research to provide additional guidance (1) to applicants on meeting the requirements of 6 61.56, (2) to the staff on evaluating applicants' methods for ensuring waste stability, (3) to better understand the long-term performance behavior of disposed wastes, and (4) to determine appropriate waste form and packaging required for above Class C wastes.
2.
Capability to assess facility engineering for disposal of LLW for deter-mining compliance with S 61.51 of 10 CFR Part 61 (1988).
Justification:
Investigation has shown that a significant path for water to enter trenches is through the trench cap.
Further, as trench cap subsidence resulting from degradation and compaction of waste packages as well as inadequate waste burial procedures and trench cap designs has shown, structural integrity is a function of the overall engineering of the disposal system, including protection of the waste form and packaging, regardless of the disposal technology used.
Research is needed to evaluate the effectiveness of various facility designs to ensure the stability of the disposal area.
3.
Capability to assess the effectiveness of monitoring plans for the preoper-ational, operational, and postoperational periods, including evaluation of the data, results, and conclusions froin the monitoring programs, to be the basis for determining compliance with applicable regulations (1988).
Justification:
Monitoring methods are needed to verify successful perform-ance of LLW disposal both during operations and following closure and to warn of incipient failure before radionuclides begin migration off site.
4.
Capability to assess the licensee's demonstration that the concentrations of radioactive material that may be releu ed from an LLW disposal facility to the general environment will meet the criteria in S 61.41 of 10 CFR Part 61, to be used in evaluating applications (1987).
Justification:
In order to determine whether the criteria in S 61.41 are met, the NRC needs to understand the methods that can be used for predict-ing radionuclide transport, including how radionuclides become available for transport from the disposal facility.
The NRC must understand the geochemical-hydrological interactions that control radionuclide transport.
This requires investigation of coupled geochemical-hydrological radio-nuclide transport models that predict transport through both saturated and unsaturated media.
8-6
5.
Capability to assess the safety of engineered enhancements and alternative technologies to SLB for disposal of low-level wastes, including wastes that have higher concentrations than are acceptable for Class C wastes, to be the basis for rulemaking, developing regulatory guides, and evaluating applications (1989).t Justification:
10 CFR Part 61 applies to all land disposal of LLW.
How-ever, technical requirements now contained in Subpart D of Part 61 apply specifically to near-surface disposal (SLB).
Disposal methods other than near-surface disposal such as aboveground vaults are likely to be presented as part of a license application.
Further, in accordance with the Low-Level Radioactive Waste Policy Amendments Act of 1985, research is needed to (1) develop performance assessment capability for licensing alternative methods that are different from conventional SLB, (2) develop an under-standing of the long-term performance of engineering materials in the disposal environment, including interaction with the LLW, that may be employed in a given alternative, and (3) establish criteria for NRC licensing of the disposal of above Class C wastes by the DOE.
6.
Criteria for characterizing sites for " alternative" LLW disposal facilities to be used in developing programs for the acquisition of site-specific data for demonstration of facility performance in compliance with NRC and EPA requirements (1990).
Justification:
Acquisition of information on a potential alternative LLW disposal site that is necessary in order to demonstrate facility perfor-mance in compliance with NRC and EPA regulations can be complicated and costly.
Not only are the data characteristic of waste retention and waste transport at the site required, but also data demonstrating compatability of the site with the engineered enhancement / alternative must be obtained.
Criteria need to be developed to allow efficient collection and analyses of relevant data with a high degree of confidence that all licensing issues will have been given adequate consideration to allow firm and consistent licensing decisions to be made.
8.2.2 Research Program Description The strategy for LLW disposal resecrch is to use field data and laboratory experiments to understand the phenomena that determine the performance of LLW land disposal facilities.
This research will be useful in guiding disposers of LLW, in assessing compliance with NRC requirements, and in evaluating the resulting level of protection achieved relative to public health and safety.
The LLW research program is directed toward supporting the regulatory require-ments of 10 CFR Part 61 and will include research addressing the disposal of wastes exceeding the limits for Class C low-level wastes.
In addition, the program studies the problems identified through experience with existing SLB for LLW disposal facilities in order to evaluate and resolve the uncertainties important to licensing.
In particular, the LLW research program is developing information that can be used to understand the factors that influence long-term trench cap stability and water infiltration through trench caps.
Also addressed is waste form degradation (which can induce trench cap failure and failure of other engineered structures).
tA modification of the regulations may result from the research.
8-7 i
\\
The LLW research program will identify means for minimizing uncertainties in predicting the release of radionuclides into the unrestricted environment. The program is also seeking to determine (1) the mechanisms that allow the release of radionuclides from the waste forms or waste packages, (2) the geochemical changes that occur when radioactive wastes interact with soils, (3) the radio-nuclides and their chemical forms that migrate through soils, and (4) the significance to waste migration of chemicals that are contained in or accompany low-level wastes.
Both testing of the chemical composition of wastes and development of data on materials that could be used to fix or retard the move-ment of radionuclides will be done.
The research will develop and test geochemical / hydrological transport models for predicting water movement and radionuclide attenuation in this water for the various media through which it may pass.
A complement to the separate parts of this program will be an effort to bring these parts together and integrate them into a systematic approach to site and facility evaluation.
This " systems integration" effort should begin in 1986 but has not yet been planned because funding projections do not indicate adequate support until FY 1987.
The major research products will be:
1.
a.
Evaluation of long-term performance of wastes and containers produced through currently available processing and containment technologies (1988).
b.
Specification of source terms from major LLW streams for assessment of site and facility performance (1989).
c.
Waste form and packaging criteria for above Class C wastes (1988).
2.
Assessment of means to enhance facility engineering performance in the control of water entry into below grade burial trenches (1987).
3.
a.
Assessment of effectiveness of monitoring methods, particularly through the use of environmental methods applicable to the post-closure period (1988).
b.
Technical basis for LLW disposal facility closure and monitoring criteria (1989).
4.
a.
Assessment of interaction of radionuclides with soils and structures to predict performance of LLW disposal facilities (1988).
b.
Description of coupled geochemical / hydrological phenomena relevant to migration of radionuclides from LLW disposal facilities (1987).
5.
a.
Assessment of safety features of engineered enhancements / alternative technologies to SLB of LLW (1988).
b.
Evaluation of appropriate technologies for licensed disposal of above Class C wastes (1990).
6.
Identification of data and information necessary to support definitive licensing decisions for an LLW site and facility (1989).
8-8
l l
l 8.2.3 Deferred /Unfunded Research Needs (as of February 1, 1986)
The systems integration, site closure and monitoring criteria, and site charac-terization criteria will not be funded until Ff 1987.
Products for source term specification and examination of alternative technologies to SLB will be provided on a delayed schedule.
Although States are actively seeking to solve their individual LLW problems, the idealized solution of a " half dozen" LLW sites may not be achieved. There may be a resultant increase in the number of sites and designs for which the present research program is insufficiently funded to pro-vide adequate and timely support to the licensing office.
Finally, work on criteria for NRC-licensed disposal of above Class C wastes by the DOE in accor-dance with the Low-Level Radioactive Waste Policy Amendments Act of 1985 will not be initiated until FY 1988.
)
8-9
Glossary ACRONYMS AND INITIALISMS ACRR Annular Core Research Reactor at Sandia National Laboratories ACRS Advisory Committee on Reactor Safeguards AE Acoustic emission ALARA As low as is reasonably achievable ASEP Accident Sequence Evaluation Program ASME American Society of Mechanical Engineers ASTM American Socf ety for Testing and Materials ATWS Anticipated transient without scram BOP Balance of plant B&W Babcock and Wilcox BWR Boiling water reactor CCTF Cylindrical Core Test Facility CFR Code of Federal Regulations CHF Critical heat flux COBRA Thermal-hydraulic code CONTAIN Containment analysis code CORA Out-of pile severe fuel damage facility in Karlsruhe, Germany CORCON Code to model interaction between molten core materials and concrete during core-melt accidents CORSOR Code to calculate core source term for severe accidents in light-water reactors G-1 i
n
DBA Design basis accident DEMONA Cuntainment model for aerosol tests at Battelle-Frankfort, Germany DESRA Code used by the U.S. Corps of Engineers to calculate seismic ground motion in loosely consolidated soils DOE Department of Energy ECC Emergency core cooling ECCS Emergency core cooling system EPA Environmental Protection Agency EPRI Electric Power Research Institute ESF Engineered safety feature FASTGRASS Code to model fission product release from fuel FIST Full Integral Simulation Test GE General Electric HDR Heissdampfreaktor (a decommissioned steam reactor in West Germany where reactor safety experiments are performed)
HECTR Code to calculate hydrogen behavior HLW High-level waste HPI High pressure injection ICAP International Code Assessment Program ICEDF Ice condenser system decontamination factor code IDCOR Industry Degraded Core Rulemaking (program)
IE (Office of) Inspection and Enforcement IEEE Institute of Electrical and Electronics Engineers IGSCC Intergranular stress corrosion cracking G-2 n
=.
i 4
INEL Idaho National Engineering Laboratory ISI Inservice inspection IST Integral System Test LACE LWR aerosol containment experiments 1
LER Lic3nsee event report LLW Low-level waste LOBI Loop blowdown investigation facility in Italy where PWR physical phenomena and parameters that affect plant performance i
during small-break accidents are studied and computer models i
developed LOCA Loss-of-coolant accident LOFT Loss-of-fluid test facility at Idaho National Engineering Laboratory i
LRRP Long-Range Research Plan LWR Light-water reactor MAAP Modular Accident Analysis Program (IDCOR's severe accident systems code)
MAEROS Code to make aerosol calculations MEDICI Code to calculate debris behavior MELCOR Code to model meltdown accident assessment (will replace MARCH, CRAC-2, and MATADOR codes)
MELPROG Melt progression code MERGE Code to calculate thermal-hydraulic parameters in detail MIST Multiloop Integral System Test MITI (Japanese) Ministry r f International Trade and Industry e
NDE Nondestructive examination NPA Nuclear plant analyzer NPAR Nuclear plant aging research l
l G-3
NPDB Nuclear plant data bank NRR (Office of) Nuclear Reactor Regulation NRU Test reactor at Chalk River, Ontario (natural uranium, heavy-water moderated and cooled)
OECD Organisation for Economic Co operation and Development OL Operating license ORNL Oak Ridge National Laboratory OTIS Once-Through Integral System OTSG Once-through steam generator PBF Power Burst Facility PISC Program for Inspection of Steel Components PMF Probable maximum flood PMH Probable maximum hurricane PNL Pacific Northwest Laboratory PPG Policy and Planning Guidance PRA Probabilistic risk assessment PTS Pressurized thermal shock PWR Pressurized water reactor QUECLA Quantitative uncertainty evaluation of containment loading analysis RAMONA Thermal-hydraulic code REIRS Radiation exposure information reporting system RELAP Detailed model for thermal-hydraulic behavior in reactor coolant system during transient and los? of-coolant accidents RES (Office of Nuclear Regulatory) Research RMIEP Risk Methodology Integration and Evaluation Program G-4
ROSA Rig of Safety Assessment (facility in Japan)
RSSMAP Reactor Safety Study Methodology Applications Program SAFT-UT Synthetic aperture focusing technique for ultrasonic testing SAR Safety analysis report SARRD Severe Accident Risk-Reduction Program SASA Severe Accident Sequence Analysis SAW Submerged arc weld SCC Stress corrosion cracking SCDAP Severe Core Damage Analysis Package SCTF Slab Core Test Facility SDMP Seismic Design Margins Program i
SLB Shallow-land burial SMAW Shielded metal arc weld SPARC Suppression pool aerosol removal code SSE Safe shutdown earthquake SSI Soil-structure interaction SSMRP Seismic Safety Margin Research Program TIC Technical Integration Center TLTA Two-Loop Test Apparatus TMI Three Mile Island TRAC Code of model core reflood and quenching TRAP-MELT Code to analyze fission product behavior within LWR primary system under accident conditions up to and including fuel melt-down UKAEA United Kingdom Atomic Energy Authority UPI Upper plenum injection G-5
UPTF Upper Plenum Test Facility USGS U.S. Geological Survey USI Unresolved safety issue VANESA Code to compute aerosol generation and fission product release during core debris interaction with concrete
-VICTORIA Code to describe fission product release from fuel and transport in reactor coolant system h
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