ML20235S279

NRC Safety Research in Support of Regulation - 1986

ML20235S279
Person / Time
Issue date:	09/30/1987
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1266, NUREG-1266-V01, NUREG-1266-V1, NUDOCS 8710080407
Download: ML20235S279 (66)

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NUREG-1266 l Vol. -1 i i

NRC Safety Research in Support of Regulation -

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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC 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 NRC 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 agencies 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.

Documerits 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 draft reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission, 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-1266 )

Vol.1 1 NRC Safety Research in Support of Regulation - 1986 i

i Manuscript Completed: April 1987 Date Published: September 1987 ,

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Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 (

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1 ABSTRACT This report is.the second in a series of annual reports responding to

. congressional inquiries as to the' utilization of nuclear regulatory research.

NUREG-1175, "NRC Safety Research in Support of Regulation," published in May 1986, reported major research accomplishments between about FY 1980 and FY.1985. This report. narrates the accomplishments of FY 1986 and does not.

restate earlier accomplishments. Earlier research results are mentioned in

.the context of current results in the interest of continuity. Both the direct contributions to scientific and technical knowledge and their regulatory-applica-tions, when there has been a definite regulatory outcome during FY 1986, have been described.

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TABLE OF CONTENTS-P_a2e.

ABSTRACT .............................................................. iii EXECUTIVE

SUMMARY

..................................................... vii h 1. SEVERE ACCIDENTS................................................. 1-1 1.I' Statement of Prob 1em........................................ 1-1 1.2 Program Strategy............................................ 1-2

1. 3 Research Accomplishments.................................... 1-2 1.3.1 Research Accomplishments During FY 1986. . . . . . . . . . . . . . 1-2 l'. 3. 2 Anticipated Accomplishments During FY 1987 and Beyond........................................... 1-3

2. RISK AND RELIABILITY............................................. 2-1 2.1 Statement of Problem........................................ 2-1 2.2 Program Strategy............................................ 2-1 2.3 Research Accomplishments.................................... 2-2 2.3.1 Reliability and Performance Indicators............... 2-2 2.3.2 Accident Frequency Estimation Methodology............ 2-3 2.3.3 Dependent Failure Methodology........................ 2-3 2.3.4 Human Reliability Analysis Methodology............... 2-4 2.3.5 Risk and Consequence Methodology...................... 2-5 2.3.6 ' Regulatory and Inspection Applications............... 2-5

3. THERMAL-HYDRAULIC TRANSIENTS..................................... 3-1

.3.1 Statement of. Problem........................................ 3-1

3. 2 Program Strategy............................................ 3-1 3.3 Research Accomplishments.................................... 3-2 3.3.1 20/3D Program........................................ 3-2 3.3.2 ECCS Rule Support.................................... 3-3 3.3.3 Experiments in Babcock and Wilcox Geometry........... 3-4 3.3.4 Experiments in Westinghouse and Combustion Engineering Geometry................................. 3-5 3.3.5 Code Development and Assessment...................... 3-7 3.3.6 Nuclear Plant Analyzer and Data Bank................. 3-8

4. PLANT AGING AND LIFE EXTENSION................................... 4-1 4.1 Statement of Problem........................................ 4-1 4.2 Program Strategy............................................ 4-1 4.3 Research Accomplishments.................................... 4-2 4.3.1 Reactor Vessels...................................... 4-2 4.3.2 Piping............................................... 4-4 4.3.3 Electrical and Mechanical Components................. 4-5 4.3.4 Nondestructive Examination........................... 4-7 v

TABLE OF CONTENTS l

l Page i 4.3.5 Equipment Qualification.............................. 4-8 1'

4.3.6 Structures........................................... 4-8

5. SEISMIC SAFETY................................................... 5-1 1

5.1 Statement of Problem........................................ 5-1 5.2 Program Strategy............................................ 5-1 5.3 Research Accomplishments.................................... 5-3 5.3.1 Earth Sciences....................................... 5-3 5.3.2 Seismic Fragility.................................... 5-3 5.3.3 Seismic Margins...................................... 5-4

5. 3. 4 Validation........................................... 5-4

6. WASTE MANAGEMENT................................................. 6-1 6.1 Statement of Problem........................................ 6-1 6.1.1 High-Level Waste..................................... 6-1 6.1.2 Low-Level Waste...................................... 6-1

6. 2 Program Strategy............................................ 6-2 6.2.1 High-Level Waste..................................... 6-2 6.2.2 Low-Level Waste...................................... 6-2 6.3 Research Accomplishments.................................... 6-3 6.3.1 High-Level Waste..................................... 6-3 6.3.2 Low-Level Waste...................................... 6-7 i

BIBLIOGRAPHY.......................................................... BI-1 GL0SSARY.............................................................. G-1 vi

EXECUTIVE

SUMMARY

j This report summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1986. This report is the second in a series of annual reports. NUREG-1175, "NRC Safety Research in Support of Regulation," published in May 1986, reported the accomplishments of the period FY 1980 to FY 1986. It is planned to continue the series of annual reports as successive volumes of NUREG-1266. The present volume is late because of events that accompanied the  !

massive NRC reorganization in April of 1987. Later volumes in the sequence should appear earlier in the calendar.

The NRC achieves its mission of providing a reasonable assurance of public health and safety from the peaceful uses of nuclear energy by maintaining  ;

multiple lines of defense against the consequences of accidents. The function ,

of research is to provide the technical basis for maintaining and strengthen- 1 ing these lines of defense, including the reduction to regulatory practice of the specific results of research.

The first line of defense is to ensure that plants are designed, built, and )

operated and maintained reliably and safely Research supporting this line of l l defense is largely concentrated within the Division of Engineering. l i

• The second pressurized thermal shock experiment, PTSE-2, was carried out j to determine if the PTS rule revision of 1985 (S 50.61 of 10 CFR Part 50) {

could be safely extrapolated to material with a low upper shelf toughness I steel characteristic of some older plants. This test confirmed the appli- {'

cability of the rule change to these older plants.

• A regulatory guide to help industry comply with the new rule was prepared as Regulatory Guide 1.154, completing the reduction to practice of the research results obtained earlier.

• Round robin eddy current tests were completed on a severely degraded steam generator. Eddy current tests are a major means of ensuring the integrity of the tubes in a steam generator. These tubes make up about 50% of the boundary between the radioactive (primary) coolant water and the nonradio-active secondary water. The results of these tests will be coupled with the results of burst testing of degraded tubes to provide a technical basis for improving the regulatory criteria for inservice inspection of steam generator tubes.

• Seismic effects on structures led to the development of a proposed revision  ;

of General Design Criterion 4, permitting the removal of counterproductive i j

snubbers from piping.

The second line of defense is to provide operating and control systems and engi-neered safety features that ensure a safe response to transient events. Work to maintain and strengthen this line of defense is focused within the Division of Reactor and Plant Systems, i

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A full-scale-test at the Upper Plenu11 Test Facility in West Germany confirmed the thermal-hydraulic conditions used as the basis for Regulatory Guide 1.154 on PTS analysis. The tests were part of the '

international 2D/3D program carried out by the NRC, Germany, and Japan.

The technical work on-the proposed revision of the ECCS rule (Appendix K to 10 CFR Part 50) was completed and summarized in a proposed revised rule forwarded to the Commission for its consideration.

e- The International Code Assessment Program (ICAP), sponsored by the NRC, developed a code uncertainty methodology. The Westinghouse Electric Corporation used this methodology in support of its submittal of the COBRA-TRAC code for NRC acceptance review.

An analysis of feed and bleed tactics to provide cooling to remove decay i heat at the Davis-Besse plant used the RELAP5 code and the nuclear plant analyser. The results indicated that if the tactics were initiated early enough, they would be successful. These results are being used in the resolution of USI A-45, " Shutdown Decay Heat Removal Requirements."

The third line of defense is-to provide 'a containment to prevent the release of dancerous amounts of radioactivity. Work to maintain and strengthen this line of cefense-is performed within the Division of Reactor Accident Analysis.

The Source Term Code Package (STCP) was released to the IAEA by Chairman Zech and made generally available. This code package had been a major part of the earlier review of-the source term work reported by the American Physical Society in Reviews of Modern Physics, July 1985, Part II.

The released package contained many revisions based on that review.

• It was established that there was no basis for a generic finding that the releases from severe accidents would be uniformly reduced in magnitude i from those projected in WASH-1400. Natural processes were identified that would, however, tend tn lead to lower predicted releases of radioactivity than those in WASH-1400 if the containment did not fail j suddenly at an early time af ter core . melt. J

• The Source Term Code Package was incorporated into the procedures used in preparing NUREG-1150, the Reactor Risk Reference Document.

A trial set of performance indicators was prepared to develop a logical method of relating observed safety performance trends at operating nuclear power plants to public risk. The process and logic models used to develop the qualitative relationship were provided by risk and reliability studies performed by RES.

• The. systems analyses of the Surry, Zion, Sequoyah, Peach Bottom, and l Grand Gulf nuclear power plants were completed in preparation for NUREG- '

1150.

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• An integrated reliability and risk analysis system was completed and implemented on a personal ccmputer to allow quick regulatory analysis of the effects om risk of system design changes, especially those that occur during maintenance operations. This code is called PRISIM.

• Human reliability research completed three projects in FY 1986: (1) a

method for appiying expert judgment in the development of generic and plant-specific human error probability estimates; (2) a computerized human reliability data bank; (3) an artificial. intelligence-based approach for modeling cognitive aspects of human behavior during normal / abnormal events n at nuclear power plants.

• The MELCOR Accident Consequence Code System, MACCS, was distributed for trial use, especially in the tracing of the consequences of the Chernobyl accident.

Waste management is a major concern of the NRC, and the same general philosophy of using multiple lines of defense applies to this concern. The research effort in support of regulation is much smaller than for nuclear power plants, however; thus, for administrative purposes the work is located within the Division of Engineering, l

• Proposed amendments to 10 CFR Part 60 were published to bring EPA high- ]

level waste standards and Part 60 into conformance. These amendments were based in large part on the technical support furnished by RES.

• Work was completed and made ready for publication on the initial phases of study of ground-water transport in saturated and unsaturated fractured rock.

• A data input guide (NUREG/CR-3162) and a theory and implementation document (NUREG/CR-3328) were prepared to aid the regulatory staff in the use of the SWIFT II code. That code contains mathematical models of the flow of ground water, transport of brine, heat, and radionuclides in saturated fractured rocks. SWIFT II is used by NMSS personnel in appraising 00E predictions of repository behavior.

• Research was concluded on sealing boreholes in crystalline rock composed of granite and basalt. The findings were published in NUREG/CR-4642 and had these major conclusions: Sealing with a swelling cement is feasible; and performance is dependent on the size of the crushed basalt and the weight ratio of bentonite to crushed basalt.

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1. SEVERE ACCIDENTS ]

1.1 Statement of Problem A severe accident in a light-water nuclear reactor can be defined as'an event in which large amounts of reactor fission products are released to the contain-ment atmosphere and potentially from there to the atmosphere outside the reactor buildings. These released radioactive fission products constitute what are called " source terms." For such an accident to happen, many low probability events must occur in succession: i.e. , some operating system must fail followed ,

by safety system failures resulting in loss of coolant, and other safety systems must fail leading to uncontrolled nuclear core overheating.

Subsequent to the Three Mile Island accident, major research efforts were under-taken by the NRC to develop an improved understanding of severe accidents and source terms in order to establish a technical basis to support regulatory decisions. The quantity, timing, and characteristics cf the release of radio-  !

active cource term materials to the environment in a severe reactor accident determine the potential hazard of the accident to the public. Because many l source term components used in the current regulations are based on technology as much as 25 years old that does not accurately reflect the best understanding. {j of the phenomena leading to radioactivity releases, this technology cannot '

reliably be used to determine how best to prevent severe accidents or mitigate their effects. Nor is this technology adequate for risk assessment.

The agency problem then is to provide a sound technical basis for taking severe accidents into account in the regulatory process in a consistent and well-formulated way. Major advances in this technology have already been made. In l

developing this technical basis, however, further research is needed to reduce uncertainties in the source term technology so as to provide better analytical tools for the regulatory evaluations. The technical issues or uncertainty L areas that have been identified for the current focus of the agency efforts are as follows:

1. Natural circulation in the reactor coolant system,

2. Core melt progression and hydrogen generation,

3. Steam explosions,

4. High pressure melt ejection,, 1 l

5. Core-concrete interactions,

6. Hydrogen ignition and burning in containment, ,

7. Iodine chemical form, and l

8. Revaporization of previously deposited fission products.

It is the characterization of the involved phenomena and the reduction of the uncertainties associated with these issues that will provide the principal focus for research efforts in connection with the development of improved severe accident analytical methodology over the next few years.

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- 1. 2 Program Strategy The NRC's strategy for taking severe accidents into account in the' regulatory process began taking shape shortly after the Three Mile Island (TMI) accident of March 1979. Approximately 175 types'of regulatory actions were proposed as a result of various studies following the accident. While most of these actions were based on information already derived from the TMI evaluation, many.were to depend on forthcoming results from the new Severe Accident Research Program.

Examples have included siting policy reformulation, analysis of hydrogen control, rulemaking on' degraded' core accidents, and improvement of the radiation protec-tion of the public.

After formulating the TMI Action Plan, the Commission developed a Severe Accident Policy Statement (July 1985) and a Safety Goal Policy Statement (August 1986).

' Major elements of the Severe Accident Policy Statement that depend on informa-tion from the Severe Accident Research Program are (1) the consideration of.

severe accidentLvulnerabilities as revealed by probabilistic risk assessments' (PRAs) for new plants, (2) the search for risk outliers by analyzing individual operating plants, and (3) the amendment of regulations, the standard review plan, and other decision procedures and criteria as needed. An implementation

. plan for the Severe Accident Policy Statement was presented to the Commissioners on February 28, 1986. This implementation plan gave further details on the three major elements mentioned above and, in particular, outlined ten potential revisions of regulatory procedures in areas such as removal of spray additives

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in PWRs, credit for fission product scrubbing in suppression pools in BWRs, emergency planning, and control room habitability.

The Commission's safety goal statement issued in August 1986 contains two qualitative safety goals that are supported by two quantitative objectives and a tentative containment performance guideline. The quantitative objectives are t'o be used in determining achievement of the safety goal. The risk analyses required by these objectives depend on the analytic capability being provided by the Severe Accident Research Program. Furthermore, the safety goal statement specifically discusses the need to treat uncertainties in phenomenological areas such as core melt progression, fission product release and transport, and containment loads and performance. Thus, the strategy for development of the detailed technical basis for evaluating severe accidents depends heavily on the careful reduction of the uncertainties mentioned in Section 1.1.

1. 3 Research Accomplishments 1.3.1 Research Accomplishments During FY 1986 The Severe Accident Research Program was largely in place by 1981 and some results were available shortly thereafter. However, a reasonably comprehensive  !

understanding of severe accidents did not emerge for 3 to 5 years. Thus, a number of years of safety research efforts have contributed to many of the regulatory applications of severe accident research that developed in FY 1986.

One of the important early results from the Severe Accident Research Program was in the area of hydrogen combustion and control. In FY 1985, results from  !

this program provided the technical bases for development of the final hydrogen control rule for PWR ice condenser containments and BWR Mark III containments 1-2

(effective February 25, 1985). In FY 1986, continuing e.ffort in this area resulted in analytical tools for licensing reviews and implementation of the hydrogen rule (published April 1986), and calculations were performed specif-ically to confirm the adequacy of the proposed deliberate ignition systems for the Sequoyah PWR ice condenser plant and the Grand Gulf BWR Mark III plant.

Also in FY 1986, data were provided by this program to close out confirmatory issues related to the hydrogen rule and to resolve the unresolved safety issue <

on hydrogen control measures and effects of hydrogen burns on safety equipment.

As a more comprehensive picture of severe accident progression began to emerge, a major effort was initiated in 1983 to reassess the technical bases for esti-mating source terms. This effort brought together all the severe accident research results then available to assemble a comprehensive (and integrated) set of computer codes for source term analysis. These codes were then used to calculate source terms for a number of important accident sequences at five selected operating plants. The integrated codes and the multiple plant calcu-lations formed the basis for an assessment of the state of the art that was issued in July 1986 as NUREG-0956, " Reassessment of the Technical Bases for Estimating Source Terms." This source term research effort provided three major practical results:

1. This effort provided a reviewed and tested set of codes for source term analysis called the Source Term Code Package. This code package was presented to the IAEA by Chairman Zech on September 26, 1986, and was made available to the general public.

2. This effort provided detailed calculations of source terms for five operating reactors using the Source Term Code Package. These detailed plant calculations were then used by the NRC as the basis for its current Reactor Risk Reference Document (see Section 2 on Risk and Reliability) l and its regulatory application. j l

3. The source term research revealed that the earlier anticipated sweeping l reductions in source terms and related reductions in the required levels f of protection were not technically justified. This has ended the drive for quick regulatory changes and placed the whole process on a more sound scientific basis.

Much of the information from the Severe Accident Research Program is being used to plan for implementing the Severe Accident Policy Statement. Two specific applications were developed in FY 1986. In one case, a standard review plan section is being revised to remove the requirement for containment spray addi-tives that were designed to remove gaseous iodine. The other is the development of a new standard review plan section to give credit for fission product removal for releases that pass through BWR suppression pools.

1.3.2 Anticipated Accomplishments During FY 1987 and Beyond Since implementation of the Severe Accident Policy Statement was just begun in FY 1986, most of the implementation activities will take place in this future period. The most important early activity in implementing this policy is the analysis of individual operating reactors in the U.S.--also referred to as the 1-3

search for risk outliers. The NRC review of the utility's individual plant analyses will be based on experimental and analytical results from the NRC's Severe Accident Research Program.

Other direct regulatory actions will also be supported by the severe accident research results. For example, Public Service Company of New Hampshire has requested a change in the emergency planning zone for the Seabrook plant.

Support from the research program is expected throughout the deliberations with the licensee on this important licensing request. Further, results from the program are essential for development of the regulatory revisions discussed in Section 1.2. Other proposed regulatory changes, such as environmental qualifi-cation of equipment and accident monitoring and management, will depend totally on information from the Severe Accident Research Program for their formulation.

Still other proposed regulatory changes, along with remaining elements of the severe accident policy, plus most elements of the safety goal, depend more broadly on risk assessments (see Section 2, " Risk and Reliability"). Keeping in mind that quantitative risk estimates come from the combination of accident frequencies (or probabilities), severe accident phenomenology (including source terms), and consequence modeling (health effects, etc.), it is clear that all regulatory applications involving risk assessments +%refore also involve severe accident issues. Results from the Severe Accid:.nt Re,earch Program will thus be an essential part of virtually all aspects of the NRC's implementation of severe accident policy and severe accident safety goals.

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2. RISK AND RELIABILITY d i!

2.1 Statement of Problem  ;

l The risk to the public health and safety presented by the current generation of j i

nuclear power plants is generally low, but for some plants the " acceptable" i

> risk level prescribed in the Commission's safety goals may be exceeded. The Commission's response to these findings must take into account large uncertain-ties in these risk estimates, the difficulty of extrapolating these results to plants for which probabilistic risk assessments have not been completed, and ,

the level of acceptable risk as defined in the Safety Goal Policy Statement. 1 The Commission's response must also reflect the results of the severe accident risk reduction study, which compares the cost and risk reduction effectiveness of a number of engineered safety systems that could be backfitted on existing reactors. This study strongly suggests that there are few, if any, backfits that could be applied to operating nuclear power plants that could reduce overall plant risk in a cost-ef fective manner.

2. 2 Program Strategy The overall NRC strategy is currently under development and will include a strengthened performance-based regulatory program of accident prevention through evaluation of operational systems, improvements in quality of construction, and standardization of new plant designs. The most effective accident prevention strategy is one that adopts a vigorous program of inspection, test and mainte-nance, and monitoring of conditions to ensure that plant safety systems will perform their intended functions with a high degree of confidence throughout the expected lifetime of the plant. In addition, reliability programs will be directed toward reducing the number of events (transients) that cause safety system actuation and toward improving operator response to abnormal events.

A risk and reliability research program to develop the additional technical information needed to support these objectives is in progress. It is divided into three elements: (1) defining the essential characteristics of competent )

and effective reliability assurance and prioritized inspection programs for nuclear power plants; (2) developing and demonstrating systems that can provide the NRC staff with quantitative measures of the success of licensee programs; and (3) providing the validated risk-based data storage and retrieval system needed to support these programs. Application of these elements will provide ,

both the NRC and the utilities with the means necessary to identity and correct j potential safety problems before they become major issues and to help to ensure that plant risk is minimized to an acceptable level. l 2-1

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2.3- Research Accomplishments 2.3.1 Reliability and Performance Indicators 2.3.1.1 Research Approach A performance indicator is a measure of the level of safety inherent in.the operation'of a nuclear power plant. Examples are the frequencies of reactor j

scrams, safety system actuations, and safety system failures. In order to '

better. monitor nuclear power plant performance and objectively determine whether plant safety is improving or degrading, the NRC has identified a set of perform-ance-indicators for initial use. While these performance indicators are quali-

, tatively related to safety, a methodology for quantitatively evaluating them from a risk.and reliability perspective is not yet available. The purpose of the performance indicators and related programs is to develop and validate such a reliability / risk-based methodology.  ;

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2.3.1.2 Research Accomplishments During FY 1986 A set of plant performance indicators was developed in 1986. These indicators were based, in part, on process and logic models developed under the NRC research 4 program. These models logically relate the performance indicators in a qualita- I tive way % public risk. NRC is using these performance indicators to help monitor safety performance trends at operating nuclear power plants.

2.3.1.3 Anticipated Accomplishments During FY 1987 and Beyond Indirect programmatic indicators reflect the effectiveness of licensee programs such as maintenance and training. Such programmatic. indicators are clearly related to safety and can be predictive in nature; however,-they are difficult to quantify on a risk basis. Research will evaluate the relationship to safety

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I of candidate indicators for maintenance, training, causes of events, operator licensing examination results, and worker radiation exposure. In addition to indirect programmatic indicators, research on direct, risk-based performance indicators will produce further improvements in NRC's capability to quantita-tively relate some of the indicators (e.g., scrams, safety system failures, and significant events) to public risk. These results will aid in monitoring plant i safety performance consistent with NRC's safety goal policy. '

The operational safety reliability research program is evaluating the effective- 1 ness of the methods and processes of reliability engineering to help achieve and maintain acceptable levels of safety. Potential applications concern the i reliability of important systems or components, such as the diesel generator, i in implementing the resolution of the station blackout unresolved safety issue and the extension of operating licenses. The operational safety reliability research program will also evaluate the effectiveness of reliability engineering methods in designing a surveillance test program that will help to assess the degree to which important safety equipment is achieving its reliability target.

The NRC will also evaluate an improved strategy for preventing multiple (i.e. ,

common cause) failures. The results of the operational safety reliability research program will help to shift regulatory requirements away from prescrip-tive requirements and focus more on safety performance.

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1 2.3.2:-Accident Frequency Estimation Methodology 2.'3.2.1 Research Approach In order to. support the NRC/IDCOR (Industry Degraded Core Rulemaking) interface on severe accident technical issues, the preparation of the Reactor Risk (

Reference Document, the Severe Accident Policy Statement, the NRC source term f reassessment, and other safety and regulatory issues, there is a need for. updated i information on the precise sequence of events in a severe accident. There is also a continuing need to evaluate probabilistic risk assessment (PRA) technology developments; integrate internal, external, and dependent failure risk methods; and identify, evaluate, and effectively display the uncertainties in PRA risk. j

-predictions.

2'.-3.2.2 Research Accomplishments During FY 1986 In support of the Reactor Risk Reference Document, the systems analysis of the 3 Surry, Zion, Sequoyah, Peach Bottom, and Grand Gulf nuclear power plants was i completed. The results from these analyses will help to assess severe core damage frequencies at nuclear power plants.

An initial version of the. integrated reliability and risk analysis system was completed. This system is an integrated PRA software tool that allows the creation and analysis of fault trees using an IBM-PC. This will aid in regula-tory evaluations of system design changes.

2.3.2.3 Anticipated Accomplishments During FY 1987 and Beyond Work on the unresolved safety issue concerning station blackout will be completed during FY 1987. This will support the final rulemaking on this important issue.

The PRA on the LaSalle plant carried out as part of the risk methods integration

, and evaluation program will be completed. The fault trees drawn for this PRA are the most comprehensive ever developed and will provide a basis for future PRA method development work. The LaSalle PRA will also provide a benchmark by which other PRAs and risk studies can be evaluated. j The PRA updating of the Surry, Zion, Sequoyah, Peach Bottom, and Grand Gulf plants will be completed. These results will form the technical basis for the Severe Accident Policy Statement.

2.3.3 Dependent Failure Methodology Dependent failures are caused by events or conditions such as environments, maintenance actions, instrument calibration, fires, floods, and seismic events ,

l affecting more than one component simultaneously or within a short period of time. Since they are major contributors to severe core damage frequency, the identification and quantification of dependent failures can significantly alter l PRA results. The strategy is to develop an integrated methodology for the identification, quantification, and assessment of the impact of dependent failures on system failures and accident sequence occurrences, i

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2.3.3.1 Research Accomplishments During FY 1986 l l'

! Screening methods for determining the most important dependent failures were developed. These methods will be tested in the LaSalle PRA under the risk-methods integration and evaluation program.

A joint effort has been initiated with the Electric Power Research Institute (EPRI) and the National Center for System Reliability of the United Kingdom to develop methods for the quantification of a class of dependent failures not explicitly modeled in the system logic models. This class of dependent failures, called common cause failures, consists of failures caused by such things as high temperature, steam, vibration, and design errors.

2.3.3.2 Anticipated Accomplishments During FY 1987 and Beyond A framework for treating dependent f ailures will be jointly developed by the NRC and EPRI. This framework will provide guidance for incorporating common cause failures into reliability studies and PRAs.

A fire risk study will be conducted to update the fire risk in four existing PRAs and evaluate the contribution and uncertainty in fire risk from six poten-tial fire risk issues: control system interaction, smoke control, fire / seismic l interaction, manual firefighting effectiveness, spurious suppression system actuation, and cable / component total environment survivability. This will provide the NRC with a better perspective on the importance of these issues and a better technical basis for evaluating the adequacy of existing regulations regarding fire.

2.3.4 Human Reliability Analysis Methodology 1

2.3.4.1 Research Approach PRAs conducted to date indicate that approximately 65 percent of all safety-related events at nuclear power plants involve human error. Methods are needed j to identify, systematically set priorities for, and suggest solutions to human i performance issues in the operation and maintenance of nuclear power plants during normal, transient, and emergency situations.

The human reliability program strategy has two objectives. The first is to develop a technical information base for human risk analysis. The second objective is to develop techniques for systematically acquiring and using qualitative and quantitative reliability and risk information to (1) realis-tically assess the contributions of human performance to overall plant risk, ,

(2) assess unresolved and generic safety issues, and (3) identify missing hunan  !

performance information. 1 2.3.4.2 Research Accomplishments During FY 1986 i

The following human reliability research products were completed during FY 1956:

(1) a method for applying expert judgment in the development of generic and I plant-specific human error probability estimates, (2) a computerized human reliability data bank for storing, aggregating, and processing human error probability data for use in PRAs, (3) an artificial intelligence-based approach 2-4 1

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, i for modeling the cognitive aspects of human behavior during no mal / abnormal events at nuclear power plants. These results were used to st4 port reliability evaluation activities of the NRC, the U.S. ndelear p wer industby, ano'the, '

international community. ,

2.3.4.3 Anticipated Accomplishn ats During FY 1987 a.nd Beyond-

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The following research projects are directed toward an improved system for conducting human reliability aslyses or to support regulatory initiatives c using risk-based data, espef 011y fcr resolving generic issues: (1) human error probabilistic data will be generatn0 Ming methods developed eactice in the program; (2) a cognitive process model based on an artificial inteViigence' '

concept from the FY 1986 program will be &veloped and field tested; a/id (3) improved procedures for integrating huma9 reliability analysis intq the FRA process will'be developed. '

s 2.3.5 Risk and Consequence Methodology The program strategy is based on the general objectives if making better data available, removing as much subjectivity K possible, an6 reducing data uncer-tainties. Major elements of the progra strategy are to (1) develop better tools to evaluate risks and consequences, (2) apply these improved calculational tools to analyze risk and consequence issues, and (3) modify appropriate nuclear i regulations in accordance with the resuits of these analyses.

2.3.5.1 Research Accomplishments During FY 198G.

TheMACCS(MELCORAccidentConsecuenceCodeSystem)andbEC(Calculationsof Reactor Acc%ent Consequences) coraputer codes were used in a number of regula-tory applications. The CRAC computer code was'used to calculate the potential '

offsite consequences of revere reactor. accidents, Both cc @r, were used_to.

estimate the risks of five operating nuclear power plante for the Reactor Risk Reference Document. The MACCS computer code was used to estimate the environ-mental consequences of the accident at Chernobyl. Both computer codes continue to be applied extensively in other regulatory areas, such as emergency planning, NRC safety goal development and application, unresolved i.afety issues, Price-Anderson legislation, and extraordinary nuclear occurrance criteria development.

2.3.5.2 Anticipated Accomplisher.ts During FY 1%7 and 3eyod MACCS will replace CRAC as the principal realistic accident consequence model in regulatory applications in the NRC. All licensing and regulatory decisio9s pursuant to the NRC's safety goal will use ret ults obtained using MACCS.

2.3.6 Regulatory and Inspection Applications There are a number of problems dealing with regulatory and inspection applica-tions of risk and reliability tchniaues.

There is a need to ascertain the safety implications of both actual and proposed plant modifications and to permit ready display of the results. The objectivt of the SARA (Systems Analysis and Risk Assessment) system is to develop a capability for computation and analysis of infortnation and plant risk ,

characteristics.

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s Theheisaneedtodevelopreliability/ risk-basedtechniquestuevaluatelicensee i submft.tals requesting modifications or exemptions from current technical specifi-cations. The objective of the PETS (Procedures for Evaluating Technical Specifi-ce Mons) and related programs is to develop and demonstrate methodologies to use reliEoility and risk techniques in evaluating the scope, detailed requirements,  ;

and safety impact of plant technical specifications.

The NRC currently has resident inspectors stationed at every licensed nuclear plant. With these inspectors making numerous inspection decisions every day, there is a need to better correlate these decisions with their potential effects on pbnt risk and for the inspectors to set priorities for their activities.

2. 3. 6.1. Research Accomplishments During FY 1986 Technical specifications are design and procedural limits that entail explicit restrictions on the operation of nuclear power plants and the maintenance of safety systems in'a pre-accident condition. Research in this area is designed to develop a methodology for setting limiting conditions for operations and j surveillance requirements for nuclear power plants based on reliability and risk analysis principles. This research now provic'es partial quantitative a

, guidance to evaluate licensee submittals requesting extensions to and/or modifi- (

l cations of teihnical specifications dealing with allowed outage times and )

surveillance : test intervals.

SARA iv a microcomputer software package that assesses the effects of such factors as component failure rates and initiating event frequencies on core melt frequency. During FY 1986 SARA was made more user friendly and was revised and expanded to provide a fairly complete basic system that permits data base review and maintenance, report generation, and graphic displays, as well as risk calculations.

PRISIM is a personal computer program that can identify and prioritize inspection activities based on their risk importance and can evaluate changes in a plant's risk status based on the operating state of the plant's systems, subsystems, and components. In FY 1986, a prototype PRISIM system was completed for the Arkansas Nuclear One (ANO-1) plant, and that system is currently undergoing field testing. 1 s

.The ANO-1 PRISIM is one of three prototype systems that will be subject to test and evaluation to establish both the system's value and its development and maintenance costs.

4 2.3.6.2 Anticipated Accomplishments During FY 1987 and Beyond t l 1

Technical guidance will be developed and provided to licensees for preparation of requests to modify technical specifications.

procedure. The development of the SARA system This will assist the NRC review will continue, using existing data pd techniques from other programs. The current SARA plant logic models

-sir be updated using results from the Reactor Risk Reference Document. SARA will be used to estimate the risk significance of observed precursor events.

By assessing the safety significance of reported operational events, SARA will foster implementation of the Commission's safety goals.

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1 PRISIM systems similar to that at ANO-1 will be developed for the Peach Bottom and Surry nuclear plants. The three prototype PRISIM systems will be installed, tested, and evaluated by the resident inspectors, regional staff, and NRC head-quarters staff with respect to their potential uses in (1) evaluating technical specifications, (2) establishing the significance of operational events at nuclear plants, and (3) assessing the risk implications of the time-varying performance of components, systems, or the entire plant.

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3. THERMAL-HYDRAULfC TRANSIENTS.

3.1 Statement of Problem The only known way to prevent reactor core melt as a result of loss of coolant is to restore coolant flow. The NRC is required to independently examine each l ,' licensee's assertions and performance of his responsibility to design, construct,

, and operate a reactor with assured coolant flow restart.

NRC's task is difficult because straightforward loss-of-coolant testing of a reactor would be too expensive and too dangerous to the public. On the other hand, straightforward and exact theoretical analyses of a reactor's fluid flows would take too long to compute because of the' complexity of heat exchange between reactor components, water, and steam, as well as of the moving mechanical inter-

. faces in pumps and the extensive baffle surfaces in the primary loops. For comparison, the much simpler, steady-state flow of aerodynamics is only now becoming computationally tractable for whole aircraft design evalunion.

1:

As a result, the NRC must combine, in a complex undertaking, limited experimental data and limited calculational capability into a firm technical basis for evalu-ating design basis accidents, ongoing events in actual reactors, and hypothetical transient threat scenarios made credible by probabilistic risk assessment and' actual events. Specific gaps in NRC information exist in the following areas:

1. Small-break loss-of-coolant accidents (LOCAs) in Babcock and Wilcox plants.

2. Evaluation of upper plenum injection, which is peculiar to six Westinghouse 2-loop plants within the United States.

3. Secondary-system transients and breaks.

4 Evaluation of BWRs without jet pumps.

5. Assessment of the NRC computer code predictions against all available experimental data.

3.2 Program Strategy The NRC has a dual, complementary approach toward achieving a firm technical I

understanding of the thermal-hydraulic behavior of the reactor. This dual approach is analytical and experimental with feedback to the analytical models.

The NRC starts by simulating the actual reactor's continuous flow of heat and fluids with a computerized model consisting of many small discrete cells exchanging heat, fluid, vapor, kinetic energy, and momentum at each small but finite time step. Physical laws are used when possible to calculate all these exchanges. Empirically derived fornMas, obtained from experiments, are used l as necessary to acccunt for such complex effects as friction between vapor and 3-1

liquid. The calculations are made for each time step and for each cell, in a l manner familiar to animated computer games, except that reactor models have many more objects (cells) and these objects interact in a tightly coupled manner at every time step, requiring many more calculations. Hence, improved efficiency in the calculations is always an objective. There are two avenues for improve-ment. First, improved numerical methods, calculating across the cells and time steps, are sought and employed. Secondly, techniques are sought to better use the hardware possibilities of vector, parallel, and distributed computing.

Our ability to rely on the computer codes for correct answers depends on three levels of experiments and comparisons of experimental results with code predic-tions. First are basic experiments to derive empirical formulas for phenomena within each cell. Next are separate effect experiments to test the code's predictions for a single, detailed component such as a steam generator. Third are integral system tests of the code predictions for a complete reactor. The results of these tests provide feedback to correct the code and our understanding of the transients. This feedback process is formalized in an international code assessment program coordinating tests and code predictions in 15 countries.

3.3 Research Accomplishments 3.3.1 2D/30 Program 3.3.1.1 Research Approach The name 2D/3D derives from the two experimental geometries, a "two-dimensional" row of fuel rod bundles and a "three-dimensional" solid cylindrical array of bundles. Each bundle is a 15 x 15 or 17 x 17 array of fuel rods. The 20/3D program is a cooperative program among the NRC, Japan Atomic Energ; W earch Institute, and the German Federal Ministry for Research and Technca gy for the study of large-break LOCAs in PWRs.

The program includes three large scale experimental facilities: Cylindrical Core Test Facility and Slab Core Test Facility located in Japan and the Upper Plenum Test Facility in West Germany.

These are the largest test facilities in the world for studying refill and i

I reflood. The NRC contribution to the program consists of advanced two phase flow instrumentation, computer codes, and analysis.

l The NRC computer model is being assessed against 20/3D experimental data to provide insights into the complex process associated with cooling the reactor core by the emergency core coolant. The general objectives are (1) to calculate the multidimensional processes during the refill and reflood stages in order to plan and coordinate the experiments and (2) to perform posttest analyses to resolve uncertainties in the model.

3.3.1.2 Research Accomplishments During FY 1986 A full-scale test at the Upper Plenum Test Facility in West Germany in April 1986 confirmed the earlier small scale test experiments in that the hot primary coolant mixes reasonably well with the incoming cold emergency core coolant in the cold legs and thus the thermal stratification and resultant stresses in the downcomer area are greatly reduced. It is estimated that the welded area in the downcomer and cold leg joints is most susceptible to pressurized thermal 3-2

i ll shock, but the thermal stratification in this area was found not to be signifi-cant. These favorable results were used to make the recommendations in Regula-tory Guide 1.154, " Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors."

A full-scale test in the Upper Plenum Test Facility in West Germany confirmed that, for typical conditions after a small-break LOCA, effective core cooling can be maintained by the flow back down the hot leg of liquid condensing from the primary steam flowing up the hot leg. This favorable result removed the concern that steam flow up the hot leg might limit the counterblow back to the vessel. These results had a major impact on the proposed revised S 50.46 of 10 CFR Part 50, the Emergency Core Cooling System Rule.

At the Cylindrical Core Test Facility in Japan, it was determined that emergency core coolant injected into the upper plenum of a reactor vessel is not blocked from the core by rising steam. Partial vertical blockages were noted but were overcome by crossflows of coolant in the core. These favorable results had two regulatory impacts. The affected plant licensees have revised their computer models to conform to test data. Of more immediate importance, if the tests had not shown successful core cooling, the upper plenum-injection plants would have needed design changes.

3.3.1.3 Anticipated Accomplishments During FY 1987 and Beyond Large-scale tests of refill and reflood phases of large-break LOCAs are scheduled for completion by FY 1989. The results will be reflected in revisions to Appendix K to 10 CFR Part 50 and in the Emergency Core Cooling System Rule.

Beneficial regulatory impact is expected on licensees seeking more economical plant operation under the revised rules.

3.3.2 ECCS Rule Support 3.3.2.1 Research Approach The research conducted since the adoption of 6 50.46 and Appendix K in 1973 has resulted in a greatly improved understanding of the phenomena that may J occur in a LOCA. With the constraints currently imposed by Appendix K, a great deal of effort expended on LOCA analysis is based on conservative assumptions that do not closely resemble the physical phenomena as they are now understood.

Moreover, distortions created by the use of those conservatism may have a net detrimental effect on the overall safety of plant design and operation. Research has been aimed at developing knowledge to allow use of best-estimate analysis, with appropriate accounting of uncertainties, rather than applying arbitrary conservatism.

Several NRC research projects mentioned elsewhere in this section contributed l to ECCS rule support. The principal projects are the performance of specific experiments in Westinghouse, Combustion Engineering, and Babcock and Wilcox test facilities; the development of analytic computer codes to scale up and integrate the test results into safety decisionmaking; and the nuclear plant analyzer software and graphics workstation to improve productivity in thermal-hydraulic analysis.

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3.3.2.2 Research Accomplishments During FY 1986 l

The research program resulting in improved understanding of ECCS performance during a LOCA has extended over a decade and was the major portion of NRC research for many years. The work in FY 1986 consisted primarily of work on the proposed revision to the ECCS rule that would allow this information to be s used in licensing. During FY 1986, the NRC staff completed work on a formal j rule revision and forwarded this proposed rule to the Commission for considera- i tion. Supporting material for this proposed rule, including a draft regulatory guide and a regulatory analysis, were also completed. It is expected that the Commission will approve publication of this proposed rule and draft regulatory guide for public comment during FY 1987. Based on the regulatory analysis performed in support of the proposed rule, it is expected that significant benefit, both safety and economic, will result from more defensible regulation making use of more realistic information gained from ECCS research.

As part of the effort to conclude this major 10 year research effort and to assist in making use of this information in licensing, a draft report was prepared to summarize the research and to serve as a guide to assist in making use of the many technical reports.

3.3.2.3 Anticipated Accomplishments During FY 1987 The proposed rule, regulatory guide, and other supporting material is being prepared for issuance for public comment in April 1987. This includes the draft NUREG-1230 Compendium of ECCS Research. After public comments are received, they will be incorporated into a final rule for issuance in early FY 1988.

3.3.3 Experiments in Babcock and Wilcox Geometry 3.3.3.1 Research Approach The small-break LOCA concerns resulting from the Three Mile Island and the more recent plant transients, such as the Davis-Besse (June 9, 1985) and Rancho Seco (December 26, 1985) events, have indicated that Babcock and Wilcox-designed ,

reactors are unusually sensitive to certain off normal secondary-side transient  !

conditions. Compared to other PWR designs, various Babcock and Wilcox design features place more reliance on the reliability and performance of the auxiliary feedwater system, emergency core cooling systems, and integrated control system to recover from transients such as loss of feedwater and loss of offsite power.

The NRC has initiated a comprehensive research program to assess Babcock and Wilcox (B&W) plant safety. This includes a short-term reassessment of plant safety for which the B&W Owners Group has the lead. NRC research is a long-term effort designed to provide the capability for improved understanding and evalua-tion of B&W plants during transients and accidents. This will allow an evalua-tion of the long-term adequacy of B&W plant design and operation.

The strategy will involve integrating the thermal-hydraulic analysis and the analysis of human factors of operator performance. In particular, several selected severe transients for a specific Babcock and Wilcox plant will be evaluated. Time-based diagrams of thermal-hydraulic analyses and operator 3-4 l

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actions will be used to assess operator performance and rec'overy procedures.

Similar transients will be analyzed on a specific Combustion Engineering plant and the findings of the different plants compared.

3.3.3.2 Research Accomplishments During FY 1986 Testing of small-break LOCAs in Babcock and Wilcox geometries was initiated in FY.1986 in the Multiloop Integral System Test facility. The results from these tests and the results of other supporting programs have been used to provide a a preliminary auessment of the adequacy of computer codes to evaluate small-break LOCAs in Babcock and Wilcox plants. Deficiencies in the RELAP5 and TRAC compu-ter codes have also been identified, and ~an overall plan for. required Sabcock and Wilcox experimental research has been completed. This research plan is now being discussed with industry in order to seek industry support in conducting this research.

Critical safety function response trees were developed for the Oconee power plant. These trees are a hierarchical representation of the resources avail-able in the plant for sustaining the critical safety functions, e.g. , core cooling. These trees were then compared to those contained in the Emergency Procedure Guidelines to. determine qualitative differences in critical safety function response in Babcock and Wilcox and Combustion Engineering plants.

Thermal-hydraulic results to date include review of the best-estimate calcula-tions for full power and a limited number of transients at reduced power levels.

Furthermore, an initial review of transients in final safety analysis reports has been completed. This will be useful to confirm safety margins in licensee analyses.

The thermal-hydraulic behavior of the Oconee plant was examined for a hypothet-

.ical transient similar to the Davis-Besse incident to assess the degree of dependence of the core cooling safety function upon the operator's performance.

This information will be useful to assess the adequacy of operator guidelines and help verify and validate probabilistic safety studies.

3.3.3.3 Anticipated Research Accomplishments During 1987 and Beyond The integrated control system model for Oconee will be completed. Critical safety function response trees for other plants will be compared to the repre-sentative Babcock and Wilcox plants for several-accident scenarios to evaluate the performance of the man and procedures together.

3.3.4 Experiments in Westinghouse and Combustion Engineering Geometry ,

3.3.4.1 Research Approach Experiments are being conducted to confirm our understanding of a spectrum of large-'and small-break transients. Particularly emphasized safety issues are those for which small changes in the physical system parameters might lead to large changes in the safety margin of conservative operation. Such effects are dif ficult to confidently calculate. Specific examples are pressurized thermal shock when emergency cooling systems operate, feed-and-bleed procedures for decay heat removal, small-break LOCAs without high pressure coolant injection, 3-5 l

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of operational transients to the point where significant boiling begins. These fast-running simulations were used for Incident Response Center training drills.

I The display _of time histories of selected plant parameters during accident l

calculations was added to the plant analyzer list of features to allow the user to assess trends in plant response upon which simulated operator action could be initiated through the interactive controls of the plant analyzer.

The nuclear plant analyzer was adapted for use with the TRAC-PF1/ MOD 1 code on the Los Alamos National Laboratory CRAY-15 computer expanding the capability of i the plant analyzer for both of the primary thermal-hydraulic accident analysis I codes in use by the NRC.

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3.3.6.3 Anticipated Accomplishments During 1987 Demonstrate nuclear plant analyzer /probabilistic risk assessment / human risk assessment methodology for in-depth analysis of loss-of-feedwater initiator in the Oconee plant, and prepare for audit of Babcock and Wilcox Owners Group recommendations for improved safety at Babcock and Wilcox plants.

Completion of the data bank program is scheduled for late FY 1987 with an operational demonstration to be held in June 1987.

More realistic training of NRC inspectors will be achieved by adding nuclear plant analyzer features to NRC's Black Fox plant simulator at the Technical Training Center in Chattanooga, Tennessee.

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4. PLANT AGING AND LIFE EXTENSION 4.1 Statement of Problem Aging affects all reactor components, systems, and structures in various degrees and has the potential to increase risk to public health and safety if its effects are not controlled. In order to ensure continuous safe operation, measures must be taken to monitor key components, systems, structures, and interfaces to detect aging degradation and to mitigate its effects through maintenance, repair, or

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replacement. For an older plant approaching the end of its design life and for which extended operation beyond the initial license period of 40 years is contem-plated, aging becomes a critical concern and will clearly be crucial to any assessment of the safety implications of license renewal.

Normal operation of a' reactor produces leakage neutrons that impinge on the walls of the reactor vessel. This action causes the steel of these walls to be embrit-tied in varying degrees depending on the location and the alloying and residual elements in the composition of the steel. Such embrittlement is manifested in a reduction of the fracture toughness of the material. Vessel walls and weldments containing copper, nickel, and phosphorus are the most susceptible to embrittle-ment. Many of the older reactor vessels in operation contain these constituents and their embrittlement increases with age.

In recent years, the nuclear industry has initiated a significant effort aimed at extending the life of existing plants beyond their original license term of 40 years. According to a.recently completed Department of Energy study, the projected net benefit to the U.S. economy can be on the order of $230 billion through the year 2030, assuming a 20 year life extension for current plants.

If a 40 year life extension is judged feasible, the benefit is even larger.

The benefit reflects both the lower fuel cost of nuclear plants compared to fossil-fueled plants and reduced outlays for replacement of generating capacity.

Utilities are currently planning to apply for license renewals and have define.:

a tentative schedule for several steps in the process. The first submittal to the NRC is expected in 1993, with a large number of additional submittals to follow shortly thereafter. To keep pace with these industry plans, the NRC will need to devote substantial effort over the next several years to license renewal. A firm NRC policy on the terms and conditions of license renewal will have to be in place by 1991. More detailed guidance for license renewal appli-cations can then be completed by early 1993. Review of these applications at I

an early stage will provide an indication to the industry of the viability of the life extension option in sufficient time to elect an alternative option if necessary.

4.2 Program Strategy The Commission in its 1986 Policy and Planning Guidance has directed that

" requests for an operating license renewal are to be anticipated." The staff l

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is to " continue development of the policies and criteria to define requirements for operating license extensions to help ensure that the industry's efforts in this area are focused on the primary regulatory concerns."

Compliance with this guidance requires NRC staff effort in several areas, including technical and scientific research to identify the effects of aging on the key safety-related components of the plant and to examine methods for miti-gating such effects. Specifically, the strategy is to achieve relative to each component the following results:

1. Identify and characterize aging and service wear effects that, if unchecked, could cause degradation of structures, components, and systems and thereby impair plant safety.

2. Develop methods of inspection, surveillance, and monitoring and of evaluating residual life of structures, components, and systems that will permit compensatory action to counter significant aging effects prior to loss of safety function.

3. Evaluate the effectiveness of storage, maintenance, repair, and replacement practices, current and proposed, in mitigating the effects and diminishing the rate and the extent of degradation caused by aging.

The program covers the reactor vessel, steam generator, piping, electrical and mechanical components and systems important to safe operation of the plant, and the containment structure. A description of the research effort as it relates to regulatory applications, both accomplished and projected, follows.

4.3 Research Accomplishments 4.3.1 Reactor Vessels 4.3.1.1 Research Accomplishments During FY 1986 l NRC-sponsored research in the earlier years had identified a potentially disrup- l tive condition for pressurized water reactor vessels that had seen considerable service. This potential condition, which is called pressurized thermal shock (PTS), can be brought about by a combination of high pressure within the vessel occurring simultaneously with a sharp reduction in the reactor coolant tempera-ture, as might happen during a small-break loss-of-coolant accident. A neces-sary condition for PTS is that the reactor vessel would have to have been in service a considerable length of time during which the fracture toughness of the vessel wall had been reduced because of radiation impingement The resolu-tion of this problem, based on preliminary research results, took the form of a 1985 amendment to the rule, S 50.61 of 10 CFR Part 50, establishing embrittlement screening criteria.

Further research was deemed necessary to validate the PTS rule. FY 1986 saw the performance of the 2nd Pressurized Thermal Shock Experiment (PTSE-2), in which material similar to the low upper shelf toughness steel found in some of 4-2

our older reactors, heated to simulate significant radiation damage, was incor-porated into an intermediate-size model test vessel and subjected to thermal and pressure transients similar to those that might occur during e severe PTS event. The results from this experiment, coupled with prior work (previously

, reported) continued to confirm that the PTS rule, S 50.61, was correct and

( slightly conservative.  !

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Significant supporting activities continued in FY 1986 at the National Bureau of Standards (NBS) to corroborate the analyses methods employed in the struc-tural evaluation L' PTS events and to establish further technical support for the PTS rule. The:s activities, which involved the testing of highly instru-mented wide plate, crack arrest experiments, using the NBS 12 million pound force test machine, developed data on the dynamics of crack formation and arrest. Data from the PTSE and wide plate series of experiments were received with great interest by the international scientific community assembled at the 2nd NRC-sponsored Crack Arrest Workshop held in April 1986. The PTS screening criteria in S 50.61 were discussed in the light of the new data presented, and there exists the possibility of its international adoption.

In another important area of study performed during FY 1986, recently obtained results indicate that the stainless steel reactor pressure vessel cladding might play a significant role in the mitigation or prevention of the formation and/or the arrest of cracks in a reactor vessel when subject to a PTS scenario.

This information was in opposition to early, preliminary work, which had appeared to indicate that the cladding would not play such a role. If further research corroborates this latter finding, then the performance of the cladding as a structural element, previously ignored in both the design stage and the later evaluation analyses of reactor vessels, could have a significant effect on the acceptance of plants for extended operation beyond the originally licensed period of 40 years.

The Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME), which is referenced in the NRC rules, has fracture toughness curves, for both crack initiation and arrest, which were used ir the establish-ment of the PTS rule and are used in the evaluation of the structural integrity of reactor pressure vessels when subjected to both normal and accident operating conditions. These curves were established on relatively sparse data dealing primarily with nuclear grade plate material. In FY 1986 the irradiation phase of the 5th Heavy Section Steel Technology (HSST) Irradiation and Fracture Toughness Test Series was completed. This experiment, which uses many specimens of several sizes up to 8 inches thick, will be completed in FY 1987 and will render data to validate the Code curves as well as verify the accuracy of our presently used predictive methods for analytically determining the irradiation-induced fracture toughness of reactor vessel steels and welds.

Regulatory application of research concentrated in FY 1986 on the development of sufficiently detailed guidance for the industry as to the ways of complying ,

with the new PTS rule. The resulting Regulatory Guide 1.154 prescribes methods i of analysis to be performed by the plant owners to justify continued operation I of the reactor beyond the screening criterion. Such justification may be based on corrective measures for reducing the rate of radiation-induced de0radation of the vessel or for mitigating the effects thereof.

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l 4.3.1.2 Anticipated Accomplishments During FY 1987 and Beyond Regulatory guidance for the industry as to the methods of compliance with the aforementioned PTS rule and the related screening criteria will be published.

This rule, the first of its kind setting reactor operating limits on reactor vessel properties, a limit beyond which a plant may not operate without the specific permission of the Commission, is being studied as a model by many nations. All testing of the 5th (present practice weld material crack initia-tion) and the 6th (present practice weld crack arrest) HSST Irradiation Series specimens will be completed in FY 1987. The results of this work will help to quantitatively define the margins of safety incorporated into the PTS rule.

One of the questions involved in the PTS rule and yet to be resolved is the margins that exist for some of the low upper shelf toughness energy welds that are found in some of our older plants. Still to be fully determined is the question as to the similarity between the irradiated performance of this material undergoing a PTS accident as compared to the present practice welds found in most of the reactors in the U.S. and as studied in the 5th and 6th HSST Irradiation Series. In the development of the PTS rule, it was assumed, based on the little evidence available, that these two weld materials would behave in a similar manner with regard to loss of fracture toughness due to irradiation. To confirm this, two new irradiation series are planned to begin in FY 1988 for completion by the end of FY 1991.

Due to the recent results indicating that the reactor pressure vessel stain-less steel cladding may be capable of preventing and/or mitigating the effect of crack formation in a reactor pressure vessel wall during a PTS event, the Third Pressurized Thermal Shock Experiment (PTSE-3) is now planned for design in FY 1988, with the experiment being conducted at the end of FY 1989 or the beginning of FY 1990, depending upon the availability of funding.

Major emphasis of this program element in the coming years will be to clearly identify and codify those factors relating to the aging degradation of reactor pressure vessels to ensure safe operation during the plant's full license period. Inherent in this work will be the identification of those parameters that might be addressed to enable the NRC to establish standards for the possible safe operational extension of the plant's license period.

4.3.2 Piping 4.3.2.1 Research Accomplishments During FY 1986 Piping design criteria research results prior to and during FY 1986 supported use of the independent support motion method in establishing licensing positions for at least two plant applications during FY 1986. Pipe damping testing and data evaluation results have supported the development of ASME Boiler and Pressure Vessel Code Case N-411 and its generic acceptance by the NRC. The Code Case, which is endorsed by the NRC, provides a new set of pipe damping criteria that serves as an alternative to earlier regulatory guidance.

Pipe cracking research results from FY 1986 included evaluation of proposed fixes for the stress corrosion cracking problem in BWRs and provided support to the technical basis of the NRC position on this prcblem, which is described in a NUREG report entitled " Materials Selection and Processing for BWR Coolant 4-4

Pressure Boundary Piping." The research also contributed to the technical basis needed to support modification of Regulatory Guide 1.44 on the control of the use of sensitized stainless steel.

Pipe fracture experiment results through FY 1986 contributed to the validation of the flaw evaluation procedures for austenitic stainless steels embodied in the ASME Boiler and Pressure Vessel Code. These data also have contributed to the development of similar flaw evaluation criteria for ferritic steels. The NRC has made use of these pipe fracture experiment results in the context of the technical basis supporting the modification of General Design Criterion 4 1 of Appendix A to 10 CFR Part 50 to eliminate dynamic effects associated with j postulated pipe ruptures. The data also have contributed to the technical basis l used in developing the " leak-before-break" acceptance criteria defined in the standard review plan.

i 4.3.2.2 Anticipated Accomplishments During FY 1987 and Beyond The piping design criteria research is expected to provide and justify recom-mended changes to the piping dynamic stress rules given in the ASME Boiler and Pressure Vessel Code. It is anticipated that completion of testing in FY 1987 will lead to major revisions of the ASME Code that will alter the dynamic design of piping systems. Ongoing research concerning the flexibility of specific piping geometries (nozzles and branch connections) is expected to result in changes to Section III of the ASME Code.

Pipe fracture research results are expected to impact the standard review plan in the context of refinement of the criteria and margins in the leak-before-break analyses. Presumably, these research efforts will lead to less restrictive criteria and reduced margins in the analyses. The associated material property research results are expected to provide data useful in plant-specific analyses and in developing flaw evaluation procedures to be incorporated in the ASME Code.

Finally, the pipe fracture research results are expected to contribute to the replacement of the postulated " double-ended guillotine break," which would lead to subsequent revision of General Design Criterion 4. This revision is expected to affect areas such as the qualification of electrical and mechanical equipment.

Consequently, significant changes would be expected in the pertinent regulatory l guides and the standard review plan.

4.3.3 Electrical and Mechanical Components l 4.3.3.1 Research Accomplishments During FY 1986 FY 1986 research concentrated on the study of the mechanisms of aging degradation and the current industry programs of inservice inspection, monitoring, and main-tenance of several components, including motor-operated and check valves, auxil-iary feedwater pumps, pipe supports (called snubbers), and emergency diesel generators. This research is part of a program to formulate effective inservice inspection, monitoring, and maintenance programs, wherever needed, that will enable informed judgment by the NRC as to where, in plant safety systems, support systems, and components, aging and degradation processes are operative and are significant to risk during normal design life and extended life. While the program is in progress, interim results are already being considered and factored 4-5

i into national standards (e.g., the ASME Operation and Maintenance Standards, IEEE), regulatory guides, and NRC information bulletins to the industry. Also,  !.

NRC plant inspection teams are being supplied with the latest advances, which equip them to perform their surveillance functions better. A few of the prom-inent accomplishments follow.

Motor-operated. valves are used extensively in nuclear power plants for service in safety-related and balance-of plant systems. Operating experience reveals numerous events in which these valves have failed to operate. A significant fraction of the events are symptomatic of aging mechanisms, and the currently required tests are not sufficient to ensure future operational readiness of the valves.

The results of FY 1986 research provided information supporting an Office of Inspection and Enforcement bulletin requesting licensees to develop and imple-ment a program to ensure that switch settings on certain safety-related valves are selected and maintained correctly.

Auxiliary feedwater pumps are critically important equipment. Failures of these pumps can reduce the amount of feedwater available for removing heat from the reactor when the usual feedwater supply is unavailable, and incidents of such failures have taken place recently. At present, such pump installations contain no monitoring devices to detect aging degradation. The research identified failure modes and causes due to aging and service wear and measurable parameters for potential use in assessing operational readiness, establishing degradation trends, and detecting incipient failures. These results are being evaluated for use in the relevant inservice inspection codes of the ASME Doiler and Pressure Vessel Code.

FY 1986 research has identified the components of the emergency diesel generators that are most susceptible to aging and are the major causes of failure. These include instrumentation and control, the fuel system, the starter, the cooling system, and the lubrication system. The results will be used to provide recem-mendations for upgrading the IEEE and ASME national standards pertaining to diesel generators.

A study was also performed in FY 1986 on the aging degradation of snubbers used as flexible restraints and supports for safety-related piping systems. In a typical commercial nuclear power plant, the number of snubbers range up to a thousand. Failures of snubbers can compromise the integrity of piping systems during thermal transients and accident situations. The results of the study on snubbers are being incorporated into a regulatory guide on the qualification and acceptance tests for sn@bers.

4.3.3.2 Anticipated Accomplishments During FY 1987 and Beyond The electric and mechanical component aging research program enters its second phase in FY 1987 when, based on the study already made, actual testing of some of the improved methods of inservice inspection and monitoring of such vital components as motor-operated valves, check valves, electric cables, and auxil-iary feedwater pumps begins.

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The program results will be used in the resolutior, of several generic safety issues, such as those pertaining to the " Adequacy of Safety-Related DC Power Supplies," "Insitu Testing of Valves," and " Diesel Reliability."

A definitive technique of diagnosing the condition of motor-operated valves--

very common, but critical, components of the reactor coolant system--based on motor current signatures, will be perfected and published in FY 1987.

Methods of pump surveillance by continaous acoustic and vibration monitoring, developed in France and Germany, are being evaluated and improved at present.

This work would produce another important tool for detecting aging degradation in critical components in time to prevent failure.

4.3.4 Nondestructive Examination 4

4.3.4.1 Research Accomplishments During FY 1986 Advances made in this area of research are reflected largely through upgrading of national standards of nondestructive examination (NDE), as contained in the ASME Boiler and Pressure Vessel Code, which is referenced in the NRC regula-tions. Research progress in FY 1986 was made in several areas that will lead to improvements in inservice inspection requirements. A few examples are cited below.

Ultrasonic inspection "round robins" and analyses of inspection team perform-l ances and equipment characteristics studies were conduc $ These were the l basic input for two ASME Code Cases on NDE system qualit e . tion requirements and criteria, including personnel training and qualification, procedure qualifi-cation and performance demonstration, and equipment qualification.

A program of round robin eddy current inspection was completed on a steam generator, which was removed from an operating plant after only 6 years of service because of serious degradation. A steam generator is a major component of pressurized water reactors,'with thousands of heat exchanger tubes, which form up to a half of tne primary reactor coolant boundary. Tubes were subse-quently removed from this generator for validation of the flaws. These results, along with results of burst testing of the degraded tubes, will form the basis for improving the current regulatory criteria for the inservice inspection for plugging of defective steam generatoi tubes, which will permit continued safe operation of the plant. ]

i During 1986, an evaluation of plant operational data and an analysis of proce-dural requirements were performed to determine the reliability and sensitivity of currently used leak detection systems and monitoring procedures. It was found that under certain conditions the present techniques and criteria may not be adequate for timely detection of small leaks and that radiation monitors are unreliable because of a high false alarm rate. The data, evaluations, and analyses developed will be used for updating the related regulatory guide and technical specifications to achieve greater reliability and sensitivity for leak detection and monitoring using current techniques.

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4.3.4.2 Anticipated Accomplishments During FY 1987 and Beyond 1

Current inservice inspection code procedures for the detection and sizing of flaws in piping and pressure vessels are not always reliable. However, an  !

improved method has been developed for the detection and sizing of flaws. Called the Synthetic Aperture Focusing Technique for Ultrasonic Testing (SAFT-UT), this method has great' promise for upgrading inspection methods for detecting flaws in austenitic stainless steel, carbon steel, and cast stainless steel piping and pressure vessels. The NRC, in cooperation with the European Organization for ,

Economic Cooperation and Development, is engaged in research to evaluate and L improve the reliability of inservice inspection.

Arrangements have been completed with the Tennessee Valley Authority for trials of continuous monitoring for structural integrity in their Watts Bar reactor during power operation using an acoustic emission technique.

4.3.5 Equipment Qualification 4.3.5.1 Research Accomplishments During FY 1986  !

Research to resolve various questions concerning the validity of procedures for the qualification testing of electric equipment has been ongoing at Sandia i National Laboratories. The results of this extensive NRC research were summar- I ized in FY 1986 in a NUREG report. This report provides guidance and recommen-dations for addressing significant questions, such as radiation dose rate effects i on aging, synergism between radiation and thermal aging, tequence of radiation and thermal aging, and accident simulation procedures.

Results of the equipment qualification testing research, betides providing guidance in the review of the licensees' qualification efforts, are being used in the development of revisions to the national standards and NRC regulatory guides. Draft regulatory guides on the " Environmental Qualification of Connec-tion Assemblies for Nuclear Power Plants," " Qualification of Lead Storage Batteries," " Electrical Penetrations Assemblies in Containment Structures for Light Water-Cooled Nuclear Power Plants" were prepared in FY 1986. These provide guidance to industry on acceptable practices for qualifying equipment.

4.3.5.2 Anticipated Accomplishments During FY 1987 and Beyond Post-Three Mile Island research on fission product release from a severe accident

, has culminated in new models for use in accident radiation source term calcula-l tions. If the radioactive release patterns predicted by the new models are

.substantially different from those currently used for equipment qualification, the relevant regulatory guide may need to be revised.

4.3.6 Structures 4.3.6.1 Research Accomplishments During FY 1986 The most severe environment considered for the design of a containment uses that of a postulated loss-of-coolant accident (LOCA) combined with the effects of a severe earthquake. Should the containment fail under such an environment, the 4-8

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consequence in terms of radioactivity release is determined most importantly by the mode and timing of failure. Early failure without other mitigating factors can result in a large radioactivity release, while delayed failure of even a few hours can significantly reduce _the amount of radioactive material available for release. Hence, the ultimate concern of the containment performance issue is how well the containment, conservatively designed for a postulated LOCA, can withstand the pressure and temperature associated with severe core damage

. accidents.

Model studies, including containment models that have been pressure-tested to destruction to validate calculational methods of predicting the behavior of real containments, are an important part of the structures research program.

Completion of the construction of a 1/6-scale model of a reinforced concrete containment for testing was a major milestone in structures research'in FY 1986.

This model, 22 feet in diameter and 37 feet high, has embedded in it several

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hundred strain gages and other sensors to study the expansion of the containment enclosure under internal pressure and temperature.

The boundary walls of a containment have openings, called penetrations, for piping, cables, and material and personnel transfer. Such penetrations are required to be perfectly leaktight, even under accident conditions. Tests on a variety of electrical containment penetrations were completed in FY 1986.

These tests form the basis for upgraded inservice inspection and testing methods to ensure their continued reliability.

4.3.6.2 Anticipated Research Accomplishments During FY 1987 and Beyond The 1/6-scale reinforced concrete containment model will be tested in FY 1987, with four European countries participating. This test will provide data that will be compared with pretest calculations, thus providing valuable insights into the mode of failure of the real containments. Even more importantly, the test results will enable the NRC to estimate with greater confidence the margin of safety available in the current containment design.

A continuation of the containment model study effort will be the testing of a model of the British Sizewell B (prestressed) containment in 1988, in which the NRC will participate.

In other efforts, a full-scale personnel transfer airlock (a penetration) will be tested to failure in FY 1987. This test will yield important information on the applicability of seals and gasket materials in the various containment i penetrations. Another series of tests will check the performance of inflatable 1 seals in containing the pressure and thermal environment of a LOCA.

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5. SEISMIC SAFETY 5.1 Statement of Problem Earthquakes are among the most severe of the natural hazards faced by nuclear power plants. Very large earthquakes can seriously challenge the safety philos- ,

ophy of redundancy and defense in depth that has been carefully designed into i the plant by creating a common mode of failure. To ensure their safety, nuclear power plants are designed to withstand the stresses imposed by such extreme external events as earthquakes, floods, tornadoes, and other violent storms.

The safety assumptions used in the plant designs are based on events that occur  ;

so rarely that the probability of the plant being faced with stresses that exceed these design bases is very difficult to quantify. As a consequence, they l have presented a major source of uncertainty in reactor design and risk assess-ment in the past. Gradually the research program has resolved the main issues with respect to meteorological phenomena as they affect reactor design and the margin of safety. Seismic hazard in Central and Eastern United States remains an issue that is not likely to be easily resolved. Historically, the largest earthquakes in the United States have occurred at New Madrid, Mo. , and at Charleston, S.C. The greater number of nuclear power plants in the Central and Eastern United States makes the seismic hazard issue one of major interest to the NRC. The problem is that the geology of the central and eastern regions makes it difficult to establish earthquake magnitudes or seismic parameters for specific locations or to ensure a proper design basis for individual power plants.

Generally, uncertair. ties such as those involved in seismic hazard analysis are not resolved quickly or in a single effort. For example, recent information from the United States Geological Survey suggests that many of the currently operating nuclear power plants, particularly in the Central and Eastern United States, may be tubjected to higher seismic loads than were specified when these plants were designed. While it is possible for a technical breakthrough to provide a definitive conclusion, it is more likely that the continuing accumu-lation of data and understanding will result in a concomitant gradual increase in the lerel of confidence held in the design basis of nuclear power plants for the Central and Eastern United States. This will require a continued level of effort in field investigation, data collection, and analysis. Faced with seismic hazard problems, the NRC needs validated analytical methods and an adequate data base to assess the capability of operating plants to sustain increased seismic loads. The research required to fill these gaps will support the development of simplified regulatory criteria and provide the NRC with an independent basis for licensing evaluations of seismic margins in operating plants.

5.2 Program Strategy The strategy to resolve the problem involves research to develop the methods and data base that will support the necessary seismic criteria development and provide the evaluation tools. The research is focused on (1) improvement of 5-1

estimates of earthquake hazards by identifying potential earthquake sources and determining the propagation of seismic energy with distance, (2) assessment of the effect of earthquakes larger than the design basis on nuclear power plant structures, systems, and components, and (3) validation of the current seismic risk analysis methods. The result will be integrated assessments of seismic safety margins at much higher levels of confidence.

The work addresses short-term, high priority research to resolve immediate licensing issues and long-term research to resolve concerns related to seismic design margins. The short-term research improves requirements by removing con-servatisms where they are unnecessary and adding conservatism where weaknesses in the regulations exist. Long-term research, on the other hand, aims at more strategic questions involving the overall view as to the importance of earth- l quakes in the regulatory process.

The program covers earth science (efforts to understand ground motion), seismic margins (capability to survive earthquakes larger than the design basis), {

seismic fragilities (seismically induced failure levels of structures or '

components), and validation (comparison of predictive methods with experimental data). With respect to earth science, the aims are to obtain a better defini-tion of the seismic hazard levels in the Eastern and Central United States, to identify the causes of seismicity, and to provide the bases for upper limits in earthquake ground motions to be used in seismic load analyses. Consistent with this is a broader objective of maintaining continued surveillance of ongoing seismic activity using seismographic networks. With respect to seismic margins, the research for reevaluation of plant seismic resistance is aimed at developing '

methods and procedures, based on the results of previous and ongoing research, to estimate the seismic margins of operating plants.

One very significant uncertainty in the seismic area is that associated with the " fragility," or failure mode and failure level, of struct ures, systems, and components. The seismic fragilities used in current licensing criteria, proba-bilistic risk assessments, and seismic margin evaluations rely heavily on subjective judgment, expert opinion, or military data. Thus, the research for resolving the uncertainties involves obtaining better estimates of the fragility of structures, components, and piping subjected to loads within and beyond their design level. In addition to providing failure data, structure research will identify how the parameters used in the design of equipment and components are affected by increased earthquake motion; component research seeks to test the hypothesis that electrical and mechanical components fail at higher levels than presently assumed and, as a consequence, that current licensing requirements are  !

overly conservative; and piping research will provide the basis for improved i balance between operating conditions in terms of overall safety.

Seismic probabilistic risk assessment methods have been developed to clarify safety issues for nuclear power plants since seismic events can simultaneously affect many plant systems and therefore be a significant or even dominant contri-butor to overall risk. tiowever, these assessment methods have not been validated to eliminate uncertainties and as such cannot yet be used with the confidence desired to make sound regulatory decisions. Research will develop sufficient l

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experimental data to allow the best possible validations of the complex probabi-listic predictive methods. Research will also evaluate the adequacy of assump-tions and subjective information used in seismic risk analyses. It also supports the seismic margins issue by providing experimental data to improve and reduce uncertainties in the current seismic design criteria.

The seismic research scope covers (in decreasing order of priority) presently operating plants, plants under construction, and possible future construction.

Disciplines covered include the geosciences and geotechnical, structural, and I mechanical engineering.

5.3 Research Accomplishments 5.3.1 Earth Sciences 5.3.1.1 Research Accomplishments During FY 1986 Seismological data from the seismographic networks on 1986 earthquakes in north-eastern Ohio aided in the resolution of the seismic design issues for the Perry Nuclear Power Plant. Fragility data compiled under the seismic design margins program was used to identify components at the plant susceptible to high-frequency input.

5.3.1.2 Anticipated Accomplishments During 1987 and Beyond The earth science research provides the data base that will be used in the proposed revision of Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants."

5.3.2 Seismic Fragility 5.3.2.] Research Accomplishments During FY 1986 A revision to a regulation based on an analysis prepared at Lawrence Livermore National Laboratory was proposed. The revision allows the elimination of arbitrary intermediate pipe restraints, and consequently the need for pipe whip restraints, in all classes of nuclear reactor piping, resulting in reduced radiation exposure to workers and improved balance on overall plant safety.

Damping tests and evaluations formed the basis for an ASME Code Case dealing with pipe damping. Implementation of this Code Case by endorsement (June 1986) permits the removal of counterproductive snubbers from piping in operating plants and plants under construction.

Seismic test data on damping and in-cabinet response amplification in motor control centers and switchgear assisted in the resolution of the unresolved safety issue on seismic qualification of electrical and mechanical components.

5.3.2.2 Anticipated Accomplishments During FY 1987 and Beyond A revision to a standard review plan eliminating the one-square-foot break area requirement in the break exclusion zone is expected in FY 1987. Only jet impingement effects will be relaxed ir the proposed revision and the associated regulatory analysis.

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Pipe rupture research will contribute to the revision of. General Design

. Criterion 4 of Appendix A to 10 CFR Part 50. This revision will avert worker radiation exposures (10,000s of person-rems) and provide cost savings of $100 million by allowing removal of certain pipe whip restraints and jet impingement shields that interfere with periodic inservice inspections.

5.3.3 Seismic Margins 5.3.3.1 Research. Accomplishments During FY 1986 High-level vibration experiments were conducted on piping components and systems to determine the actual mechanisms of failure. The results of these experiments will be used to improve the ASME Boiler and Pressure Vessel Code design criteria for piping, resulting in more flexible, safer pipe designs. . Analytical studies of the independent support motion method were used to establish the current regulatory position on the use of this spectral analysis technique for piping.

Steel containment buildings could buckle under seismic loads. Research conducted on this behavior has produced an improved ability to predict such buckling char-acteristics. NRC research on buckling behavior has been used to verify the adequacy of ASME Code design rules in this area, which were found to be adequate.

i 5.3.3.2 Anticipated Accomplishments During FY 1987 and Beyond The review methodology developed under the seismic design margins program will be used in implementing the Severe Accident Policy Statement and in assessing specific plant seismic resistance for earthquakes above current design levels.

For example, the result of a trial seismic margins review of the Maine Yankee plant will be used in 1987 as direct input in a safety evaluation report on seismic issues for that plant.

Results from research on reinforced concrete buildings will provide the technical bases for cost-benefit assessments and revisions to regulatory guides on damping values for the seismic design of nuclear power plants, on combining modal responses and spatial components in seismic response analysis, and on the development of floor design response spectra for the seismic design of floor-supported equipment or components. Revisions will also be made to standard review plans on seismic design parameters, seismic system analysis, and seismic subsystem analysis.

5.3.4 Validation l 5.3.4.1 Research Accomplishments During FY 1986 l

In 1986, research results have been used in resolving an unresolved safety issue on seismic design criteria. They have also been used to make proposed changes to the standard review plan section relating to soil-structure interactions.

5.3.4.2 Anticipated Accomplishments During FY 1987 and Beyond Ongoing joint efforts with the Electric Power Research Institute and with West Germany are expected to produce validation of the calculational methods used for licensing evaluations to predict component responses to seismic loadings.

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6. WASTE MANAGEMENT 6.1 -Statement of Problem ,j 6.1.1 High-Level Waste The waste management issue involves both high-level waste and low-level waste.

The high-level waste (HLW) disposal policy for the United States is defined by 4 the Atomic Energy Act, the Energy Reorganization Act, and the Nuclear Waste Policy Act. The latter, signed into law in 1983, provides for the development of geologic repositories for the permanent disposal of high-level radioactive

~ waste and assigns, responsibility for repository development to the Department

-of Energy (DOE), environmental standards development to the Environmental Protection Agency, and regulation to protect public health and safety and the environment to the NRC. Among its many provisions, the Nuclear Waste Policy Act established schedules and funding mechanisms for repository development.

Although the majority of actions required under the Act are to be performed by the DOE, the Act assigned to NRC the responsibility of licensing these facili-ties. The Act directs DOE to identify and characterize three sites by 1990.

At the present time, 00E has indicated to Congress that they cannot meet the 1990 date and is now planning onsite characterization by 1992. -At that time, DOE is scheduled to choose a site and submit a license application for construc-tion authorization to the NRC.

An HLW repository poses unique considerations and uncertainties related to waste emplacement, monitoring, and performance assessment. Much of this unique-ness stems.from the type of facility, a first-of-its-kind geologic disposal installation, and the fact that it will be placed in low permeability / low flow geologic systems that have not been investigated previously because of their low economic value. NRC must have an independent capability to evaluate DOE .,

safety analyses and predict whether long-term releases will be within established limits. The NRC research program objective is to provide the technical capabil-ity necessary to evaluate DOE's site characterization activities and to assess DOE's license application when it'is submitted.

i 6.1.2 Low-Level Waste i Disposal of low-level waste (LLW) involves issues at the forefront of technology, for example, waste form, waste package integrity, and long-term retention of radionuclides in the disposal facility environment. Research is required to establish regulatory criteria to permit sound evaluation of proposals for disposal facilities and to ensure that all regulatory requirements, particularly  ;

those on radionuclides release limits, will be met. Establishing these criteria in a timely manner is made more urgent and complex by two factors. First, the Low Level Radioactive Waste Policy Amendments Act of 1985 (P.L.99-240) set a very tight time schedule for establishing facilities within the various States.

Second, the States and compacts of States have chosen to consider alternative disposal methods to shallow land burial. Certain of these alternatives must be uitically examined by tightly focused research to determine their acceptability and to give guidance to the States.

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The direction of the LLW research program has responded to legislative mandates I that resulted from an earlier history of shallow land burial of wastes at a i number of sites for several decades that used vague and differing criteria as  !

to site suitability, waste package design, etc. Disposal criteria for LLW have evolved as experience, knowledge, public awareness, and political controversy have grown. In particular, through the Low Level Radioactive Waste Policy Amendments Act of 1985, the Congress has required the NRC to quickly complete the development of riound technical bases for regulatory decisionmaking regarding engineered LLW disposal methods, so-called alternative LLW disposal. This change has broadened the scope of NRC LLW research. ,

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6.2 Program Strat$gy 6.2.1 High-Level Waste The research program has been guided to provide the technical foundation for NRC development of a set of regulations for the review and licensing of HLW reposi-tories. This freework for NRC review will allow the formal licensing activities and the supporting research to be focused on significant issues.

At present, the NRC has active research programs in hydrology, geology, materials science, geochemistry, and several other disciplines related to the management l of high-level waste. The research combines theoretical study with laboratory and field experiments to identify the physical processes that determine repository performance in the types of genlogic media found at sites currently under consideration by 00E. Theultime.tegoaloftheNRC'swastevanagementresearch is to provide the technical basis to support the licensing staff s independent judgment as to the appropriateness and adequacy of DOE's demonstration of compliance with 10 CFR Part 60 and the EPA's HLW standard. In addition, NRC's waste management research seeks to provide technical support to the licensing ,

staff in their interactions with 00E and the States and to develop regulatory ,/

standards to support the licensing of facilities and methods for the disposal c .

and management of high-level radioactive wastes.

6.2.2 Low-Level Waste At present NRC research in support of licensing activities for LLW disposal facilities is focused on addressing (1) water entry into disposal units,

' (2) performance of waste forms and waste packages, (3) characterization of the LLW source term, (4) mecharIsms for transport of radionuclides from the disposal units, and (5) the safety and performance of engineered enhancements and alter-natives to conventional shallow land burial for LLW disposal. This research is intended to support not only the NRC licensing but also those States that regu-late 1.LW disposal. This diverse ussr community makes the coordination and

, . definition of LLW research and the dissemination of products a much more compli-cated undertaking than that for the HLW program. Further, the fact that many States will be the licensors and are looking to the NRC for technical support in their licensing and regulatory programs drives the LLW research to be more prescriptive and developmental than is the HLW research program.

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6. 3 Research Accomplishments I i 3 6.3.1 High-Level Waste s

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6. 3.1.1 ? esearch R ApprcNc.h y 7 .

The NRC has active research programs in hydrology, p ology, materials l science, geochemistry, and seveqal 3ther discfalines related tc the management of high-I

., j, level waste (HLW). The research' combines theoretical .tudy with laboratory and field experiments to ident,ify the physical processes that control and determine iI e'

repository performance in the tyqes of geologic media found at sites currently under consideration by.00E. Thi ultimate goal of the NRC's waste manag,'taant research is to provide the technidal basis for the licensing staff to make its .-

own independent judgment as 'to %e appropriateness and adequacy of DOE's demon

• stration of compliance with 10 CFP, Part 60 and the EPA's HLW standard. In this  :

regard, the most mmediately usable regulatory application of the HLW research j program is the ongoing technical support to the licensing staff in the b continuing prelicensiag discussions with the 00E. In the following narratWe, specific'results of NRC HLW researchproduced over the last fiscal year will be described, beginning with the most visible regulatory application-- 1 rulemakings.

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6.3.1.2 Research Accomplishments During FY 1986 ..

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1. Rulemakings ,

In June 1986, proposed amendments to Part 69 to eliminate inconsistencies between the EPA HLW standards and Part 60 t ere published for comment. NRC has received a wide variety of public comments that are being considered in the preparation of the final amendments.' r In July 1986, the NRC published final prucedural amendments to Part 60 dealing with site characterization and the participation of States and 3 Indian tribes in the licensing process for an HLW repository. These amendments were needect to bring the procedures in Part 60 into ' conformity 5 with those established by the Nuclear Waste Policy Act of 198l'.'

Results of HLW research conducted by NRC/RES has formed the major portion of the technical tasa for 10 CFR Part 60, as well as support for the abcve ,y, changes to Part 60. #

2. G_eohydrology t j

Since transport by ground water is the most likely path by which radio- ,

active nuclides from the disposed waste can reach the environment, the J NRC is actively studying the movement of ofrund water in the types of media being considered by D0E. Experimentii sites have been located in fractured rick, both above and below tt'e water tablu, and field tssting is oeis con &ci.cd to determine what type of measurements are appropriate to <+,arccterize the hydrology of f ractured media and how measurement, data should be analyzed to MMC ground-water flow. A field study in saurated fractured rock was inituted in September 1985 to test the relationships between field measurement df parameters and model scales derived from earlier work. The importance of larle natural anomalous hydrologic

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.' S j features,: appropriateness of continuum versus discrete fracture models, measurement of effective porosity, theories of spatially projecting dispersitivity measurements, and distinction between matrix diffusion, dis.ersion,'and o sorption are among the questions that are being addressed in this study. The initial phases of a similar study examining unsaturated rock were completed in August 1986. This work focused on assessing techni-quep and methodologies for fracture characterization, infiltration, and percolation studies, rock and matrix permeability tcsting, vapor phase flow and transport assessment, and numerical simulations of flow and trans-port.;in partially saturated media. Results are being published in NUREG/

CR-4659 Volumes 1 and 2. In addition, the NRC, the University of Arizona, and Sandia National Laboratories jointly sponsored a special workshop on t " Unsaturated Flow and Contaminant Transport Related to High-Level Radio-activo Waste Disposal," in January 1986. Processes and field studies dealing with unsaturated flow and solute transport phenomena were discussed, and laboratory and field data were presented.

1 l Also, in this past year, a data input guide (NUREG/CR-3162) and a theory and iniplementation document (NUREG/CR-3328) were issued associated with 4

SWIFT II, a computer program developed by NRC's waste management research program, which contains mathematical models of the flow of ground water, e r transport of brine, transport of heat, and transport of radionuclides in  ;

l saturated fractured rocks. The development of SWIFT II has provided the j

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L , licensing staff with a tool by which they can make a direct independent l.4 comprism and confirmation of DOE predictions of HLW repository behavior l'I,O #

over very long times.

l . $ Investigating the performance that can be expected from the waste form and

• 3 waste package is another major area of NRC's HLW research. NRC-sponsored q research programs are identifying and studying the mechanisms of waste I package and waste form failure under expected repository conditions. These

) studies are essential if NRC is to be able to independently evaluate DOE's l demonstration that the waste form and waste package comply with the contain-l ment and controlled release requirements of 10 CFR Part 60. During 1986, i the corrosion research groups under contract to NRC used statistical methods of experimer .il design, combined with cyclic voltammetry, to assess the potential for failure of HLW metal overpacks by stress corrosion cracking and localized corrosion, As a result of this work, a significant new understanding of localized corrosion in carbon steel, of particular I I

importance to geologic disposal of HLW, was realized.

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' New work on the properties of spent reactor fuel as an HLW waste form was begun this year, and the Japan Atomic Energy Research Institute (JAERI) began a series of experiments of waste package and HLW glass waste form performance in high radiation environments as part of the NRC-JAERI research information exchange agreement. j

3. Geochemistry s

The NRC has an active research program in the vital field of geochemistry '

'related to the management of HLW. During this past year, researchers demonstrated that radionuclides solubilities predicted at 60 C, using presently accepted models and thermodynamic data measured at 25*C, do not

. compare well with actual solubilities measured at 60 C (NUREG/CR-4582).

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New solid phases for which 25 C data were unavailable appeared to be controlling solubilities at 60 C. These results supported the licensing ,

staff discussions with DOE as to the need for DOE to develop thermodynamic data over a realistic range of temperatures to understand repository radionuclides' release.

In NRC-sponsored studies of bentonite packing and backfill materials for use_in HLW repositories, it was found that surface diffusion increased transport ~through bentonite significantly over what had been expected.

As a consequence of this work, NRC now has a more realistic assessment of the bulk diffusion coefficient appropriate for use by 00E in modeling the performance of bentonite packing and backfill material.

A mathematical model of radionuclides transport using laboratory-based measurement was used to predict the distributions of radioisotopes being transported away from a uranium ore body. The predicted movement agreed very well with actual field observations. This work, conducted in coopera-tion with the Australian Atomic Energy Commission at a uranium ore site in the Northwest Territory of Australia, has provided the licensing staff l with an improved understanding of the nature and significance of the correspondence between performance predictions based on laboratory obser-vations and what is likely to be actual repository behavior.

4. Borehole and Shaft Sealing i NRC's research program to assess experimentally the performance of existing technology for sealing boreholes at HLW repositories is continuing. During 1986 sealing research in crystalline host rock composed of granite and j basalt was concluded. Some of the significant findings, reported in -

NUREG/CR-4642, were as follows:

• Sealing horizontal-boreholes in the field with a swelling cement appears quite feasible.

• The hydraulic conductivity of cement plugs installed in a borehole increases about two orders of magnitude when the temperature is raised from about ambient (22 C) to about 90-95 C.

The size of a cementitious borehole plug affects its sealing capability.

• Cement grout distribution in rock fractures was found to be uneven. l However, the grout reduced the hydraulic conductivity of the fracture.

• The performance of bentonite / crushed basalt borehole plugs is dependent on the size of the crushed basalt and the ratio of bentonite to crushed basalt, by weight.

These results will be used by the licensing staff in their review of DOE's program to provide effective sealing of shafts and boreholes at a reposi-tory site, 6-5  ;

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6.3,1.3 Anticipated Research Accomplishments in FY 1987 and Beyond

1. Rulemakings Final amendments to 10 CFR Part 60 conforming NRC's requirements to EPA's I HLW standard and an advanced notice of proposed rulemaking to clarify the definition of HLW in accordance with the Nuclear Waste Policy Act are the HLW rulemakings scheduled for FY 1987. The NRC's HLW re. search program will provide technical support to these rulemakings as required.

2. Specific Technical Topics Research results anticipated over the next fes years are cited below. The specific use of a research product in the future HLW licensing process is not yet well defincd. However, anticipated results can be grouped accord-ing to the specific performance objectives of the NRC's HLW regulation, 10 CFR Part 60, addressed by this research. ,

a. Evaluation of Releases to Accessible Environment Identification and description of important parameters and functional l relationships for hydrological and geochemical models for assessing radionuclides transport processes (1988).

Report on field validation studies of radionuclides transport model (1990).

Parameters and processes important to evaluation of effectiveness of borehole plugging and sealing and shaft sealing techniques for salt (1988).

Parameters and processes important to evaluation of effects of cyclical re-wetting in tuff on borehole and shaft seals (1989).

Assessment of effects of likely off-normal conditions that might result in oegraded repository performance (1990).

b. Evaluation of Containment Requirement Identification and description of the relationship among parameters important in container manufacturing and expected long-term container performance (1987).

Identification and description of methods for predicting long-term j performance of waste packages in tuff (including waste form, container, and overpack) (1989).

Identification and description of waste package performance in the environment of a salt repository (1990).

c. Evaluation of Release Rate Requirement on Engineered Facility Paramete rt, and physical relationships important to methods for evaluating j long-term performance of backfill systems proposed by DOE (1988).

6-6

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Applicability of hydrothermal and geothermal data and predictive techniques to waste isolation performance assessments (1987).

Identification and description of important parameters and functional relationships for hydrological and geochemical models for assessing radionuclides transport processes (1989).

6.3.2 Low-Level Waste 6.3.2.1 Research Approach Technical information is required for evaluating and licensing LLW disposal facilities. Research is needed to provide guidance and data for making licensing and inspection decisions for developing needed regulations on LLW and packages and for assessing engineering and burial enhancements to shallow land burial for low-level waste. In May 1983, the NRC issued a technical position paper that specified minimum performance standards for LLW waste forms. NRC will need to conduct research to evaluate.the longevity of the packaging of LLW as required by 10 CFR Part 61. In addition, the chemical and radioactive properties of decontamination wastes from operating nuclear power stations need to be characterized to establish realistic source term data in modeling the migration of radionuclides at LLW disposal sites.

The NRC staff needs to understand the uncertainties and information needs of various models and computer codes presently used and anticipated to be used by DOE and their contractors in performance assessment studies. An understanding of hydrogeologic models and their accompanying technical assumptions is needed to ensure that regulatory decisions are based on adequate information.

NRC research in support of licensing activities for low-level radioactive waste (LLW) disposal facilities is focused on (1) water entry into disposal units, (2) performance of waste forms and waste packages, (3) characterization of the LLW source term, (4) mechanisms for transport of radionuclides from the disposal units, and (5) the safety and performance of engineered enhancements and alterna-tives to conventional shallow land burial for LLW disposal. This research will be useful not only to the NRC licensing staff but also to States regulating LLW disposal.

6.3.2.2 Research Accomplishments During FY 1986 The NRC's LLW research program has responded to the combination of legislation dealing with disposal of LLW (Low-Level Radioactive Waste Policy Act and the Low-Level Radioactive Waste Policy Amendments Act), the development and promul-gation of the Commission's LLW regulations, 10 CFR Part 61, and the present hiatus in development of new LLW disposal facilities by refocusing the emphasis onto technical issues that are or will be dealt with by States and State compacts in the development and regulation of new LLW disposal facilities. As a result i of this refocusing which began about 3 years ago, research products are only now beginning to emerge from the NRC's LLW research program. For this reason, most results obtained in FY 1986 represent first steps rather than finished products.

6-7

-1. ~ Alternatives to Shallow Land Burial There is great interest on the part of States and State compacts in alter-natives to shallow land burial as it is currently practiced. Work was completed' to' identify and assess the importance of the key engineering .

design and safety features of a number of alternatives being considered by States. This work has identified for licensing evaluation some engineered features common to many alternatives, as well as the relative contribution to safe disposal made by various features within a given alternative.

'2. LLW Source Terms Taking advantage of the waste classification and waste form and packaging

' requirements (10 CFR Part 61), NRC is engaged in research to develop one or more model LLW source terms. These source terms will be usable as input to LLW disposa1' performance assessment models as part of the LLW licensing review process. The exact nature and number of source term development will evolve from findings as to t6e' degree of chemical and j physical variability of the dominant waste radionuclides occurring in wastes classified and packaged in conformance with 10 CFR Part 61.

3. Radionuclides Transport An NRC-sponsored cooperative project between Atomic Energy of Canada Ltd.

(AECL) and the Pacific Northwest Laboratories (PNL) has been using data collocted from 40 years of LLW disposal at AECL's Chalk River facility to assess techniques for modeling LLW site performance. PNL is approaching

.the problem as though dealing with a pristine site, prior to waste disposal.

They have been going through various stages of site analysis and site review, using those portions of the data base required to resolve licensing issues.

The full data base from over 120 wells is being used to assess the validity of the model predictions and analyses based on a small data set typical of what a license applicant would be expected to develop during site charac-terization. Results to date indicate that such a limited data set (16 wells in this case) might be expected to conservatively bound the transport plume of radionuclides. Further, predictions based on the complete data set of 120 existing wells at the Nitrate Disposal Area did not offer signif-icantly improved results. This exercise lends confidence to the practicality l of modeling a site using a well-chosen date tet collected during site characterization. The project is now proceeding to look at a more complex site at Chalk River to confirm the results of this earlier work at a relatively simple site. This project is providing important insights into the design of data evaluation programs for future LLW disposal and the reliability of predictions based on the data. This project should be completed in FY 1987.

4. Hydrology In August 1986, a report was issued to present the experimental and modeling results that were used in the design of a field experiment to obtain data for validating the stochastic flow and transport models. Initially, the field work will focus on stochastic models. Ultimately, the study results 6-8

will be used to provide guidance to the States and licensees for LLW site characterization and performance assessments.

6.3.2.3 Anticipated Research Accomplishments During FY 1987 and Beyond

1. Infiltration of Water Evaluation of a bioengineered system to control water entry through trench covers (see figure) was begun. Preliminary results indicate that the combination of engineering and vegetation appears to be very resistant to failure from cover subsidence or deterioration of the material on the cover because it effectively controls deep water percolation beneath the cover. 3 l

The results of this work will be applicable to any disposal scheme employ-ing earthen covers. A report presenting an assessment of this means to enhance control of waste entry into LLW disposal facilities is anticipated in FY 1988.

2. Alternatives to Shallow Land Burial Work is continuing on the assessment of the reliability and expected longevity in disposal environments of the engineered components. The results of this work are expected to be available by January 1988.

3. Radionuclides Transport Researchers are looking at the role played by vegetation in radionuclides migration. They have found that plant roots exude agents that tend to mobilize radionuclides to a degree greater than previously anticipated.

The results of this research are being factored into geochemical / hydrologic transport models that will be used for predicting the performance of an LLW disposal site. A report is expected in FY 1987.

The Office of Nuclear Regulatory Research has a contract with the National Research Council to evaluate techniques for estimating probabilities of extreme floods. A panel of experts assembled by the Water Science and Technology Board of the National Research Council has held three meetings since January 1986 to review methods of determining extreme flood probabil-ities and make recommendations to the NRC. A report covering major portions of their report summarizing the state of the science and recommending interim methods and further research has been prepared. Completion of the final report is anticipated in early 1987.

4. LLW Waste Forms and Waste Packages In May 1983, the NRC issued a technical position paper that specified ,

minimum performance standards for LLW waste forms. Current waste forms  !

in commercial use were tested under contract to NRC to ensure that leaching characteristics and compressive strength of waste forms are consistent with the standards specified in the technical position. Various decon-tamination wastes from actual power plants using commercial solidification processes such as Lomi, Candecon, NS-1, and Citrox are being investigated.

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- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ . /

RES also is beginning a program to develop testing protocols for demon-strating expected lifetimes for so-called high integrity containers for LLW disposal. Such protocols will facilitate NRC's evaluation of the longevity of the packaging of LLW as required by 10 CFR Part 61.

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6-11

BIBLIOGRAPHY

'American Society of Mechanical Engineers (ASME), " Boiler and Pressure Vessel

-Code,"Section III.

Bonzon, L. L., et al., " Status Report on Equipment Qualification Issues Research and Resolution," Sandia National Laboratories, NUREG/CR-4303, SAND 85-1309,

' November 1986.

Hazelton, W. S., " Technical Report on Material Selection and Processing Guide-lines for BWR Coolant Pressure Boundary Piping," NUREG-0313, Rev. 2, June 1986.  ;

Idaho National Engineering Laboratory (INEL), " User's Manual for USNRC's Nuclear Plant Analyzer," EG&G Idaho Report EGG-RST-7044, September 1985.

INEL, " Integrated Reliability and Risk Analysis Users Guide (IRRAS-PC),"

October 1986.

Kirchner, J. R. , and D. J. Campbell, "An Overview of the Plant Risk Status Information Management (PRISIM)," Fourteenth Water Reactor Safety Information Meeting, NUREG/CP-0082, Vol. 1, p. 179, February 1987.

Krantz, E. A., et al., " System Analysis and Risk Assessment (SARA) System,"

Fourteenth Water Reactor Safety Information Meeting, NUREG/CP-0082, Vol. 1,

p. 193, February 1987.

National Research Council of the National Academy of Sciences, "An Assessment

, of Techniques for Estimating Probabilities of Extreme Floods," second draft, February 1987.

. Reeves, M., et al., " Data Input Guide for SWIFT II - The Sandia Waste-Isolation Flow and Transport Model for Fractured Media Release 4.84," Sandia National Laboratories, NUREG/CR-3162, SAND 83-0242, April 1986.

Reeves, M., et al., " Theory and Implementation for SWIFT II - The Sandia Waste-Isolation Flow and Transport Model for Fractured Media Release 4.84," Sandia National Laboratories, NUREG/CR-3328, SAND 83-1159, August 1986.

Ritchie, L. T., et al., " Calculations of Reactor Accident Consequences, Version 2, CRAC2 Computer Code: User's Guide," Sandia National Laboratories, NUREG/CR-2326, SAN 081-1994, April 1983.

Ritchie, L. T., et al., "CRAC2 Model Description," Sandia National Laboratories, NUREG/CR-2552, SAND 82-0342, April 1984.

Samanta, P. K., et al., " Procedures for Evaluating Technical Specifications (PETS)," Fourteenth Water Reactor Safety Information Meeting, NUREG/CP-0082, Vol.1, p.163, February 1987.

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Silberberg, M., et al., " Reassessment of the Technical-Bases for Estimating Source Terms," NUREG-0956, July 1986.

Taylor, D. D. , et al. , " TRAC-BD1/ MOD 1: An Advanced 8est Estimate Computer _

Program for Boiling Water Reactor Transient Analysis," EG&G, Inc.,

NUREG/CR-3633, Vols. 1-3, EGG-2294, April 1984.

'U.S. Nuclear Regulatory Commission (USNRC), " Reactor Safety Study--An Assess-ment of Accident Risks in U. S. Commercial Nuclear Power Plants," WASH-1400 (NUREG-75/014), October 1975.

USNRC, " Policy Statement on Severe Reactor Accidents," Federal Register, Vol. 50, p. 32138, August 8, 1985.

USNRC, "U.S. Nuclear Regulatory Commission Policy and Planning Guidance 1986,"

NUREG-0885, Issue 5, February 1986.

USNRC, " Policy Statement on Safety Goals for the Operation of Nuclear Power Plants," Federal Register, Vol. 51, p. 28044, August 4, 1986.

USNRC, " Reactor Risk Reference Document," NUREG-1150, Vols. 1-3, Draft Report I for Comment, February 1987.

USNRC, "NRC Action Plan Developed as a Result of TMI-2 Accident," NUREG-0660, May 1980.

USNRC, " Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports- for Pressurized Water Reactors," Regulatory Guide 1.154.

USNRC, " Compendium of ECCS Research for Realistic LOCA Analysis," draft NUREG-1230, April 1987.

USNRC, " TRAC /PF1/ MOD 1: An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Thermal-Hydraulic Analysis," Los Alamos National Laboratory, NUREG/CR-3858, LA-10157-MS, July 1986.

I USNRC, " International Code Assessment and Application Program (ICAP),"

NUREG-1270, Vol. 1, March 1987.

USNRC, " COBRA-TRAC: A Thermal Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems," NUREG/CR-3046, Vols. 1-5, March 1983.

USNRC, " Unresolved Safety Issues Summary: Data as of August 16, 1985,"

NUREG-0606, Vol. 7, No. 3, August 1985. I USNRC, " Standard Review Plan for the Review of Safety Analysis Reports for I Nuclear Power Plants--Light Water Reactor Edition," NUREG-0800, July 1981.

USNRC, " Qualification and Acceptance Tests for Snubbers Used in Systems i Important to Safety," Draf t Regulatory Guide SC 708-4.  ;

USNRC, " Environmental Qualification of Connection Assemblies for Nuclear Power Plants," Draft Regulatory Guide EE 404-4.

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4 USNRC,. " Electrical ' Penetration ' Assemblies in Containment Structures' for Nuclear Power Plants," Regulatory Guide 1.63.

USNRC, Office of Inspection and Enforcement, " Masonry Wall Design,"

Bulletin 80-11, May 8, 1980.

USNRC, " Damping Values for' Seismic Design of Nuclear Power Plants," Regulatory Guide 1.61.

USNRC, " Combining Modal Responses and Spatial Components in Seismic Response Analysis," Regulatory Guide 1.92.

USNRC, " Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components," Regulatory Guide 1.122.

USNRC, Office of Inspection and Enforcement, " Motor Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch-Settings,"

Bulletin 85-03, November 15, 1985.

Wierenga, P. J., et al., " Validation of Stochastic Flow and Transport Models for Unsaturated Soils: A Comprehensive Field Study," Pacific Northwest Laboratories, NUREG/CR-4622, PNL-5875, August 1986.

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GLOSSARY AEC Atomic Energy Commission. The independent civilian agency of the Federal government that had statutory responsibility for atomic energy matters prior to the creation of the Nuclear Regulatory Commission.

Aerosol A suspension of particles contained in a gas. When the particles are very small, they can remain suspended for a long time (hours).

Anticipated An improbable nuclear power plant occurrence that Transient Without assumes complete failure of the plant to scram (see Scram (ATWS) Scram) when scram is callea for and proceeds until other action is taken either automatically or by an operator.

ASME American Society of Mechanical Engineers. A professional society that develops standards in many engineering areas, including the nuclear area.  ;

\ \

BWR Boiling water reactor. A nuclear reactor in which i water is used as both a coolant and a moderator. It is allowed to boil in the core, and the resulting steam is used directly to drive a turbine and then condensed and returned to the reactor vessel.

10 CFR Title 10 to the Code of Federal Regulations. Title 10,

" Energy," is composed of four volumes. The first volume, Parts 0-199, contains the regulations of the Nuclear Regulatory Commission.

Cold Leg The input pipe for primary coolant returning from the steam generator to the reactor vessel.

I Containment An enclosure around a nuclear reactor to retain {

radioactive materials that otherwise might be released 1 to the atmosphere in the event of an accident.  !

Control Rod A rod, plate, or tube containing a strong neutron absorbing material (hafnium, boron, cadmium, etc.)

used to establish or eliminate nuclear criticality. A control rod absorbs neutrons, preventing them from causing further fissions. j Coolant A fluid that is circulated through a nuclear reactor to remove heat. Water is used in most U.S. power reactors. f<

I L

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~l; Core! The central portion of a nuclear reactor containing the fuel. elements, moderator, neutron poisons, and J support structures; the volume where fission occurs,  !

' producing heat.

]

Core Melt .This term applies to the' overheating of a reactor core as a result of the failure of reactor shutdown or cooling' systems,. leading to substantial melting of the radioactive fuel'and the structures that hold the fuel ,

in place. I Critical Heat Flux- The heat flux above which the nucleate boiling process

-(CHF) breaks down and trans'ition boiling begins. For heat transfer from a wall to a liquid, increasing the wall-

• liquid temperature difference will result in decreased heat transfer to the liquid because of vapor formation along the wall-liquid interface. The maximum heat transfer attainable per unit area-time is called CHF.

Decommissioning Steps taken after permanent shutdown of a reactor to  !

ensure continued safety and noncontamination of the environment.

l' Defense in Depth .. An engineering practice involving multiple types of protection against accidents. It includes quality .;

assurance and control in plant design, construction, '

y and operation; backup systems; engineered safety features to confine the consequences of accidents; siting; and emergency planning.

Deflagration A combustion wave that is traveling at a speed that is subsonic relative-to the unburned gas. Most fires are of this type.

Design Basis A ' requirement that limits the design of a facility, i system, or item of equipment.

Design Basis Accident The most serious reactor accident for which the ,

system is designed. .]

Detonation A combustion wave that is traveling at a speed that is supersonic relative to the unburned gas. Associated with an explosion.

Direct Containment If a pressurized water reactor vessel is ruptured in Heating (DCH) a severe accident, molten core debris could be ejected under high pressure into the surrounding containment.

The resulting high pressure melt ejection could frag-ment and react with oxygen and steam, thus generating chemical energy that produces heat and pressurizes the containment atmosphere, and could produce hydrogen that may burn and further pressurize the containment.

G-2

s Downcomer The annular space between the core and the vessel side walls through which incoming primary coolant is directed to the bottom'of the reactor vessel.

ECCS Emergency core cooling system. A backup system designed to keep the nuclear fuel cooled in the event that normal coolant is lost.

EPRI Electric Power Research Institute. A research organi-zation owned and operated by a consortium of U.S.

electric power utilities. EPRI conducts nuclear and non-nuclear research.

Equipment Process by which the capability of equipment to survive Qualification and function during design basis events is demonstrated. 4 l

Failure The inability of a system or component to perform its intended function.

Fission The splitting of a heavy nucleus into two or more parts (which are nuclei of lighter elements) accompanied by the release of a relatively large amount of energy and frequently two or more neutrons. Fission can occur spontaneously, but usually it is caused by the absorp-tion of neutrons, gamma rays, or other particles.

Fission Products Nuclei formed by the fission of heavy elements. Almost all are radioactive.

Fuel Fissionable material used, or usable, to produce energy in a reactor. The term is also applied to a mixture such as natural uranium, in which only part of the 3 atoms are readily fissionable, provided the mixture <

can be made to sustain a chain reaction.  :

Fuel Cladding The outer jacket of nuclear fuel elements. It provides structural support of the fuel and prevents the release of fission products into the coolant. Aluminum, stain-less steel, and zirconium alloys are typical cladding materials. Zirconium is the currently preferred material. ,

HLW High-level waste (see radioactive waste).

IDCOR Industry degraded core rulemaking program.

IGSCC Intergranular stress-corrosion cracking. Corrosion cracking in BWR piping caused by high local stress, sensitization of material, and high oxygen content in water. The corrosion occurs along the boundaries between different grains.

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1 Instr'ument Tubes The metal tubes containing sensors that enter'the reactor pressure vessel and core region from the vessel bottom.  :

LLW- Low-level waste (see radioactive waste).

LOCA Loss-of-coolant accident. A LOCA can result from breaks in pipes carrying the coolant or leaks from valves that control its flow.

LWR Light water reactor. A nuclear reactor in which ordinary water is used as the coolant. PWRs and BWRs are both LWRs. l NRC Nuclear Regulatory Commission. The Federal agency.

responsible for the regulating, licensing, and inspecting of nuclear facilities and materials and for conducting the research necessary to perform these functions.

NRR (Office of) Nuclear Reactor Regulation, NRC. The entity within the NRC . responsible for licensing nuclear reactors.

Nuclide A species of atom that can be distinguished by.its atomic weight, atomic number, and energy state. A L

radionuclides is a radioactive nuclide.

Performance Indicator A measure of the level of safety inherent in the (PI) operation of a nuclear power plant, such as the frequency of scrams.

Pipe Whip Severe movement of broken piping arising from reaction

.to water flow changes, similar to an uncontrolled <

garden hose whipping around.

Pressure Vessel A strong-walled container that houses the core of most' power reactors.

Pressurized A phenomenon of cracking under conditions of rapid Thermal Shock cooling at high pressure (which may occur in some accident sequences); it may occur in certain pressure vessels embrittled from years of neutron radiation, l depending on the material composition.

PWR Pressurized water reactor. A nuclear reactor in which  ;

heat is transferred from the core to a steam generator i by water kept under high pressure to achieve high i temperature but prevent boiling in the core.

Probabilistic Risk The art of quantifying a risk based on estimated Assessment or component and human failure rates and the anticipated Probabilistic Risk consequences associated with these failures, which may ,

Analysis (PRA) occur either singly or in combination. Probabilistic  !

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risk assessment typically involves the use'of event trees and fault trees, although these are not the only tools available for such assessments.

Radiation. The propagation of energy through matter or space in the form of waves. Includes particles (alpha and beta rays, free neutrons, etc.) and electromagnetic radia-tion (gamma and x-rays).

Radiation Dosimetry The measurement of the amount of radiation delivered to or absorbed at a specific place. '

Radioactive Waste Equipment and materials (from nuclear operations) that are radioactive and that are being discarded. Wastes are generally referred to as high level, having radio-activity concentrations of hundreds to thousands of curies per gallon or cubic foot such as spent nuclear fuel, or low level, in the range of 1 microcurie per gallon or cubic foot, which can include contaminated clothing or medical wastes.

Rogulatory Guide An NRC publication that is used to describe and make available to the public methods acceptable to the NRC staff of implementing specific parts-of the Commission's regulations, to delineate techniques used by the staff in evaluating specific problems or postulated accidents, and otherwise to provide guidance to applicants. Regu-latory guides are not NRC requirements.

RES (Office'of) Nuclear Regulatory Research, NRC. The entity within NRC responsible for conducting research needed to support the regulatory process.

Risk The product of the probability of occurrence and the consequence. Both the probability of occurrence and i the consequence of a release to the environment are l important for assessing nuclear plant safety.

Scram The sudden shutdown of a nuclear reactor, usually by l rapid insertion of safety rods and control rods. Emer- '

gencies or deviations from normal reactor operation cause the reactor operator or automatic control equipment to trip the reactor.

Seismic Pertaining to an earthquake.

Seismicity Relationship of the frequency and distribution of earthquakes.

SEP Systematic Evaluation Program. A program in which eight of the older operating plants were evaluated against the intent of the current licensing criteria for selected issues.

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Severe Accident An incident in a nuclear-reactor that continues to the point that cladding is damaged and fission products are released from the fuel rods. Radioactivity may or may not be released outside the reactor containment.

For example, the incident at Three Mile Island-Unit 2 was a severe accident. Severe accidents exceed design basis accidents.

Snubber A shock absorber attached to piping to lessen the effects of vibration or' movement.

Source Term The quantity, timing, and characteristics of the release of radioactive material to the environment following a postulated severe reactor accident.

Steam Generator A boiler assembly composed of many tubes where secondary turbine water on the outside of the tubes is heated to steam by hot reactor primary coolant on the inside of ]

the tubes. Steam generators are used in PWRs.

i- Technical A design or procedural restriction on the operation of Specification a nuclear power plant or the maintenance of a safety 0 l system in a pre-accident condition.

Thermal-Hydraulic Pertaining to fluid movement and heat transfer within the reactor system.

Transient A malfunction or deviation from normal reactor operation.

Upper Plenum The space inside the' reactor vessel above the core region.

Waste Form The radioactive waste material and any encapsulating or stabilizing matrix, such as spent reactor 002 fuel and borosilicate glass.

Waste Package The waste form and any other containers, shielding, packing, overpacking, and other absorbent materials immediately surrounding an individual waste container.

Water Hammer A physical phenomenon, associated with a loud banging noise, that occurs when water flows are suddenly I interrupted. Water hammers often exert large forces on pipes and are accounted for in reactor designs.

Water Impingement The impact of water or steam under high pressure striking another piece of equipment following a pipe break.

1 I

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