Design-Basis Accident Analysis Methods for Light-Water Nuclear Power Plants

Robert P Martin, Cesare Frepoli

• Publisher: WSPC
• eBook ISBN: 9789813275676
• Topic: Technology & Engineering
• Subtopic: Mechanical Engineering
• Index: Technology & Engineering

This book captures the principles of safety evaluation as practiced in the regulated light-water reactor nuclear industry, as established and stabilized over the last 30 years. It is expected to serve both the current industry and those planning for th

CHAPTER 1

Regulatory Status

S. M. Bajorek

1.1Introduction

The International Atomic Energy Agency (IAEA) defines nuclear safety as “The achievement of proper operating conditions, prevention of accidents or mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation hazards”. In the interest of promoting and ensuring a high level of safety for nuclear projects, a national nuclear safety authority or regulatory body serves the public interest through the administration of regulations commensurate with particular uses of nuclear materials. Authorization of nuclear-related activities and facilities through a process of licensing is one of the principal functions of the nuclear regulator.

While nuclear regulation addresses broad issues, nuclear power plant safety analyses involving design-basis transients and accidents uniquely intersect technical, commercial and public policy issues. In particular, it is viewed by regulators as an essential part of the verification of a design with respect to the plant and its engineered safety features. Such engineered safety features are expected to be resilient to credible events, while those associated with normal operation and, to a lesser extent, to a few exceptional events that are not expected to occur over the plant’s lifetime, are examined to demonstrate design robustness. This chapter introduces the nuclear regulatory framework specific to the preparation and review of design-basis accident analysis methods for light-water nuclear power plants.

1.1.1Role and history of the regulatory authority

The regulator has a unique role in technology development and research, and this role is unique by design. The origins of this tradition began with the Atomic Energy Act of 1946 in the United States, which transferred control of nuclear technology from military to civilian control. The Act established the U.S. Atomic Energy Commission (AEC) with the mandate to foster and control the peacetime development of atomic science. The Act, which created the AEC, transferred many nuclear laboratories that were instrumental in the initial understanding of nuclear science and safety, from military to civilian control. It also put the AEC in the politically difficult “dual-role” position of simultaneously ensuring safety to the public and also of promoting what eventually grew into the nuclear power industry. This dual role received increasing criticism during the 1960s and in 1974 the AEC was replaced by the Energy Research and Development Administration (ERDA) and the U.S. Nuclear Regulatory Commission (NRC). Several years later, ERDA was combined with several other agencies associated with energy production into what is now the U.S. Department of Energy (DOE).

With the 1974 reorganization, the role of the NRC was made clear. The NRC was to be independent and responsible for public health and safety. As part of its mission, the NRC is responsible for reactor licensing and renewal, licensing of radioactive materials, spent fuel management and safety, and security. The DOE is responsible for a wide range of energy-related areas, including nuclear energy-related research, energy production and waste disposal. While some areas appear to overlap, the goal of the reorganization was to establish the DOE as a supporter of the (nuclear) industry and the NRC as the independent regulator.

1.1.2Regulatory requirements

The U.S. Code of Federal Regulations, Title 10 — Energy, contains the regulations governing the civilian use of radioactive material. While, of course, only enforceable in the United States, much of the regulations have generally been adopted by all other nuclear power nations. The regulations appearing in its “Part 50” or 10 CFR 50 [U.S. Nuclear Regulatory Commission, 2017a] apply explicitly to licensing power reactors. In historical context, the regulations in 10 CFR Part 50 “are promulgated by the Nuclear Regulatory Commission pursuant to the Atomic Energy Act of 1954, as amended and . . . the Energy Reorganization Action of 1974, to provide for the licensing of production and utilization facilities”, which include nuclear reactors.

1.1.2.1Regulatory criteria

The regulatory criteria governing design-basis accident analysis can be viewed hierarchically beginning with rules addressing the principal nuclear safety concern: uncontrolled radiological releases. At this top tier are standards for protection against ionizing radiation that address the dose limits and other requirements required for a comprehensive radiation safety program, which appear in 10 CFR Part 20 [U.S. Nuclear Regulatory Commission, 2017b]. Closely associated with a radiation safety program is plant siting. 10 CFR Part 100 [U.S. Nuclear Regulatory Commission, 2017c] addresses plant siting which explicitly considers the reactor design, the population density near the site and the site’s physical characteristics. It includes the acceptance criteria, in terms of the radiation dose, that are ultimately quantified from interpretations of design-basis accident analysis.

In defining the term “Safety-Related Structure”, 10 CFR 50.2 states that design-basis events are conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events and natural phenomena for which the plant must be designed to ensure:

•The integrity of the reactor coolant pressure boundary;

•The capability to shut down the reactor and maintain it in a safe shutdown condition; or

•The capability to prevent or mitigate the consequences of accidents which could result in significant offsite exposures.

While many of the regulations are specific to light water reactors, 10 CFR 50.43(e) applies to all nuclear reactors. It provides guidance on design certification applications regarding the demonstration of safety system performance and data sufficiency for assessment analytical tools. The performance of each safety feature must be demonstrated through either analysis, test programs and experience, or combination thereof. Applications for design certification under 10 CFR Part 52 [U.S. Nuclear Regulatory Commission, 2017d] must identify interdependent effects involving the safety features and provide sufficient experimental data to assess analytical tools for safety analysis. The data must span a sufficient range covering normal conditions and transient conditions, in addition to accident sequences.

10 CFR 50.43(e) also provides an alternative approach to licensing. A design can be found acceptable if there has been sufficient testing performed with a prototype plant, and that testing covers a range including normal, transient and accident conditions. If the prototype is used to comply with the testing requirements the NRC may elect to impose additional requirements on plant siting, safety features or operational conditions.

The acceptance criteria for emergency core cooling systems (ECCS) in light-water reactors are defined in 10 CFR 50.46. The regulation is specific to light-water reactors with zirconium-clad fuel elements and is directed towards the analysis of loss-of-coolant accidents (LOCAs). While specific with regard to reactor and fuel design and a LOCA, it is often used as a de facto guide for other designs and accident scenarios. The regulatory criteria set forth by 10 CFR50.46 are:

(1)The calculated maximum fuel element cladding temperature must not exceed 2,200 degrees °F (1,204°C).

(2)The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3)The total amount of hydrogen generated from the chemical reaction of the cladding with water or steam must not exceed 0.01 times the hypothetical amount, which would be generated if all of the cladding material were to have reacted.

(4)Changes in the core geometry must be such that the core remains coolable.

(5)After initial operation of the ECCS, the core temperature must be maintained at an acceptably low level, demonstrating the capability of long-term cooling.

10 CFR 50.46 allows for two alternative approaches to calculate the ECCS performance: the use of 10 CFR 50 Appendix K [U.S. Nuclear Regulatory Commission, 2000] or use of a realistic analytical technique. Appendix K, generally considered a highly conservative approach, defines modeling techniques and acceptable features of an evaluation model. Some notable requirements of Appendix K include:

(a)The heat generation rates from radioactive decay of fission products shall be assumed to be equal to 1.2 times the values for the infinite operating time in [American Nuclear Society, 1971].

(b)The steady state temperature distribution and stored energy in the fuel before the hypothetical accident shall be calculated for the burn-up, which yields the highest calculated cladding temperature (or, optionally, the highest calculated stored energy).

(c)It must be assumed the reactor has been operating continuously at a power level of at least 1.02 times the licensed power level (to allow for instrumentation error), with the maximum peaking factor allowed by the technical specifications.

(d)The rate of energy release, hydrogen generation and cladding oxidation from the metal/water reaction shall be calculated using the Baker–Just equation [Baker and Just, 1962].

(e)For all times after the discharging fluid has been calculated to be two-phase in composition, the discharge rate shall be calculated by use of the Moody model [Moody, 1965].

(f)For postulated cold leg breaks, all emergency cooling water injected into the inlet lines or the reactor vessel during the bypass period shall, in the calculations, be subtracted from the reactor vessel calculated inventory.

(g)An analysis of possible failure modes of ECCS equipment and of their effects on ECCS performance must be made. In carrying out the accident evaluation, the combination of ECCS subsystems assumed to be operative shall be those available after the most damaging single failure of ECCS equipment has taken place.

(h)During refill and during reflood when reflood rates are less than one inch per second, heat transfer calculations shall be based on the assumption cooling is only by steam, and shall take into account any flow blockage calculated to occur as a result of cladding swelling or rupture, as such blockage might affect both local steam flow and heat transfer.

(i)During the period following lower plenum flashing but prior to the core spray reaching rated flow, a convective heat transfer coefficient of zero shall be applied to all fuel rods.

These are not a complete list of Appendix K restrictions, but are the ones most likely to result in conservatism in a LOCA analysis. Since its publication, however, experimental efforts have been able to quantify the conservatism and a revision to 10 CFR 50.46 was enabled.

1.1.2.21988 BELOCA rule change

After extensive rulemaking hearings, 10 CFR 50.46 and Appendix K were formally published in January 1974, providing a guide to the evaluation of an ECCS. An evaluation model was defined as the computer code, or system of codes, and calculations used to determine the effectiveness of the ECCS to mitigate a loss-of-coolant accident (LOCA). Because technology was limited and there were few experimental studies available, the rule and Appendix K restrictions were conservative. At the time of enactment, the state of the art for analysis of a LOCA was in its infancy and the Appendix K restrictions on analysis methods were a reflection of the uncertainties in simulating many of the physical phenomena in a LOCA. A close examination of the Commissioner’s Opinions [U.S. Atomic Energy Commission, 1973], however, and a later study by the American Physical Society (APS) [Lewis et al., 1975] on the recently published rule left no doubt that the rule was probably conservative. However, the complexity of the evaluation models and uncertainty in various elements made it difficult to demonstrate this conservatism. The APS study group proposed an approach to establish the degree of conservatism associated with system designs that were similar to the proposed approach in the 1988 revision to 50.46:

“. . .If the degree of conservatism of the ECCS...

Table of contents

• Cover Page
• Title Page
• Copyright Page
• Preface
• About the Authors
• Acknowledgments
• List of Tables
• List of Figures
• 1. Regulatory Status
• 2. The Safety Case
• 3. Design-Basis Event Characterization
• 4. Analytical Requirements and Software
• 5. Verification and Validation
• 6. Similarity and Scaling
• 7. Deterministic and Best-Estimate Analysis Methods
• 8. PWR LOCA/Non-LOCA Design-Basis Events
• 9. BWR LOCA/Non-LOCA Design-Basis Events
• 10. LWR — Reactivity Transients and Accidents
• 11. LWR Impact on Containment
• 12. Radiological Evaluations
• 13. Accident Tolerant Designs and Corresponding Analyses — Generation IV/SMRs
• 14. Licensing Considerations
• Glossary
• Index