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# Nuclear Reactor Safety Research

On this page:
        - Reactor Fuel Behavior and High Burnup Fuel
        - Plant Aging
        - Plant Material Conditions
        - Digital Instrumentation and Controls
        - Thermal Hydraulic and Severe Accident Computer Codes
        - Probabilistic Risk Analysis
        - Operational Data Assessment
        - Radiation Protection
        - Human Performance
        - State-of-the-Art Reactor Consequence Analyses

## Reactor Fuel Behavior and High Burnup Fuel

Postulated accidents are studied to demonstrate that fuel damage during such events would be adequately limited and to confirm that recent increases in energy output or the use of new cladding alloys in the fuel rods preserve the limits on fuel damage.

## Plant Aging

A better understanding of the effect of age-related degradation on structures, systems, and components is being developed to ensure that adequate margins are maintained under all design conditions for the current and any extended operating life of nuclear power plants. The research is currently focused on the age-related degradation of passive components such as cables, connectors, and penetrations.

## Plant Material Conditions

This research is focused on measuring, evaluating, and predicting the effects on structural integrity of nuclear systems, structures, and components (e.g., reactor pressure vessel and internals, and steam generator tubes) during normal operations. The research also focused on a range of postulated accidents during which the plant experiences hostile environments that include high temperatures, stresses, nuclear irradiation, aggressive water coolant chemistry, cyclic loading, and general wear. The research incorporates both laboratory and field measurements, evaluations, and predictions to enhance the agency?s technical basis for regulation.

## Digital Instrumentation and Controls

Because digital instrumentation and controls (I&C) are easier to obtain than analog I&C and because digital controls can easily perform more complex functions, plants are expected to retrofit their safety, operating, and support systems with digital I&C. The research is focused on evaluating these changes to digital controls, their possible safety implications, and their complexity and unique failure modes to provide timely guidance to regulators and the industry based on the research findings.

## Thermal Hydraulic and Severe Accident Computer Codes

Thermal-hydraulic research activities at NRC focus on the development of computer codes that simulate the behavior of the reactor system to ensure that a balance between energy removal from the fuel to the coolant is balanced by energy production in the fuel. The computer codes are used to assess the consequences if an imbalance occurs and to determine the effectiveness of mitigating actions. NRC maintains experimental research programs to obtain data that are used in both computer code assessment and model development activities. Non-proprietary experimental data and associated code input decks can be made available to the public upon request.

Severe accidents are the highly improbable group of accidents that involve serious, prolonged overheating of most of the nuclear fuel that then could result in the release of large amounts of radiation and radioactive material. Research studies of severe accidents assess the detailed behavior of reactor and containment systems, including the means by which these accidents may be prevented or mitigated. Research studies address fuel damage, progression of accident scenarios, ability to maintain damaged fuel within the reactor pressure vessel and, in the event of reactor vessel failure, the ability to confine the radiation release within the containment building.

## Probabilistic Risk Analysis

The NRC's 1995 Probabilistic Risk Assessment (PRA) Policy Statement states that "the use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data." NRC's PRA research program contributes to the implementation of this policy in that it applies current methods and data to the resolution of reactor safety issues and develops and demonstrates new state-of-the-art PRA methods.

## Operational Data Assessment

The NRC research program has activities that collect, analyze, and disseminate operational data and assess trends in performance. These data and assessments provide insights to improve the understanding of the risk significance of events that have occurred at licensed facilities. Examination of individual plants and events at these plants is used to develop guidance on the use of risk assessments for the agency's reactor oversight program.

## Radiation Protection

The radiation protection research program collects, analyzes, and disseminates information on occupational exposures reported to NRC by licensees. This information is used to track the effectiveness of licensee As Low As Reasonably Achievable (ALARA) programs and will form the basis for future studies to evaluate the health effects of this group of workers.

## Human Performance

The Human Performance Research program focuses on the interaction of people with the systems and the environments in which they work. It establishes the technical basis for NRC initiatives in areas such as inspection guidance for evaluating emergency operating procedures, a systems approach to training, human system interface design for current and advanced control station design, human performance contributors in events, communications-related corrective action plans, shift working hours, and fatigue management programs.

## State-of-the-Art Reactor Consequence Analyses

The State-of-the-Art Reactor Consequence Analyses (SOARCA) program involves the reanalysis of severe accident consequences to develop a body of knowledge regarding the realistic outcomes of severe reactor accidents. In addition to incorporating the results of more than 25 years of research, the objective of this updated plant analysis is to include the significant plant safety improvements and updates, which have been made by plant owners but were not reflected in earlier assessments by the U.S. Nuclear Regulatory Commission (NRC). In particular, these plant safety improvements include system enhancements, training and emergency procedures, and offsite emergency response. In addition, these improvements include the recent enhancements in connection with security-related events.

Page Last Reviewed/Updated Tuesday, July 7, 2020

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          - Nuclear Reactor Safety Research
          - Nuclear Materials Safety Research
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          - Accident Sequence Precursor (ASP) Program
          - The Radiation Protection Computer Code Analysis and Maintenance Program (RAMP)
          - The Code Application and Maintenance Program (CAMP)
          - Digital Instrumentation and Controls (I&C) Research
          - State-of-the-Art Reactor Consequence Analyses (SOARCA)
          - Computer Codes
          - Obtaining the Codes
          - Navigating Space Nuclear Safety: The Role of the Nuclear Regulatory Commission
          - Analysis of Cancer Risk in Populations Near Nuclear Facilities

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Page Last Reviewed/Updated Tuesday, July 7, 2020