This page is under construction
Considerations in the development of safety requirements for Nuclear Technology Modular Pebble Bed Reactor
SECTION ONE: INTRODUCTION TO THE SAFETY CASE OF NTPBMR
The Founder and Principal Engineer of AscenTrust, LLC. has been working on the design of the Nuclear Technology Modular Pebble Bed Reactor (NTPBMR) for many years. Our prospective Joint Venture partner has requested us to assess, at the conceptual stage, the safety of the design of our pebble bed modular nuclear reactor concept. The NTPBMR is part of a class of advanced reactor technologies loosely referred to as Generation IV reactor technologies. A copy of the findings of the Generation IV is available from the Senior Engineer. These advanced reactors rely on a variety of technologies and are of a high degree of innovation. However, to date, for the Pebble Bed Modular Reactor and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist.
This document is an outcome of the efforts expended by the Engineering Staff of AscenTrust, LLC. to develop a general approach for assessing the safety of the design of the pebble bed reactor, and of reactors in general, with characteristics that differ from those of light water reactors. This document addresses, specifically, the problem of the safety case for the NTPBMR. The safety assessment procedure to be used for our nuclear prototype are based on the well-established and accepted principle of defense in depth. There is a need to develop a general approach for assessing the safety of the design of our reactor technology and other reactor technologies which do not involve water as a coolant or a moderator. This need to evaluate the safety case applies to all kinds of advanced reactor technologies.
The NTPBMR reactor technology, in common with other modular high temperature gas cooled reactors adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors.
This document is the result of the development of the method proposed. The approach presented is based on an extended interpretation of the concept of defense in depth and its link with the general safety objectives and fundamental safety functions as set out in our “Safety of Nuclear Technologies Power Plants: Design”. The present document is not intended to be exhaustive, but rather suggests a systematic approach to be used by Nuclear Technology, Inc. in the development of detailed safety requirements for the NTPBMR prior to our application for a special license to build a prototype of the NTPBMR technology. The prototype will be operated to provide the data required to develop licensing procedures and a new set of CFR’s formulated specifically for the Pebble Bed Reactor Technology.
Gas cooled reactors have had a long and varied history which dates back to the very early days of the development of nuclear energy. One of the most successful types of gas-cooled reactors has been the Pebble Bed Modular Reactor. Document 00130, a technical document draw up by AscenTrust, LLC. is a compilation of information on the status of the design and safety for gas cooled reactors. The evolutionary process, along with significant advances in supporting technologies, have culminated in the Nuclear Technology pebble bed modular reactor (NTPBMR). The NTPBMR is expected to achieve the goals of safe, efficient, environmentally acceptable and economic production of energy at high temperature for the generation of electricity and for industrial process heat applications in the twenty-first century.
The NTPBMR concept originated in Germany in the early 1960’s. There were parallel design variations in the USA and other countries during the 1960s and early 19700s but the essential systems and sub-systems of the NTPBMR were mainly derived from the reactor which was built and operated at the Juelich Research Center, in Germany. This 15 MWe demonstration reactor, Arbeitsgemeinschaft Versuchsreaktor (AVR–roughly translated to working-group research reactor or working-group experimental reactor), was built at the Julich Research Center, in Julich West Germany. The goal was to gain operational experience with a high-temperature, pebble-bed, graphite moderated and gas-cooled reactor. The unit’s first criticality was on August 26, 1966. The facility ran successfully for 22 years, and was decommissioned on December 1, 1988 in the wake of the Chernobyl disaster.
AVR pebble bed research reactor rated at 40 Mwth and 15 Mwe, operated for 22 years, demonstrating that this technology works. The reactor produced heat by passing helium gas through a reactor core consisting of uranium fueled pebbles. The heat was transferred from the helium loop through a heat exchanger to a steam cycle and the conventional steam cycle was used to generate electricity through a conventional steam electric plant. Germany also built a 300 Mwe version of the pebble bed reactor but it suffered some early mechanical and political problems that eventually lead to its shutdown.
In the Netherlands, the Petten Research Institute is developing pebble bed reactors for industrial applications in the range of 15 MWth.
In December of 2000, the Institute of Nuclear Energy Technology of Tsinghua University in Beijing China, achieved first criticality of their 10 MWth pebble bed research reactor. In China, they are still operating this 10 megawatt thermal, 4 megawatt electric pebble bed reactor that is now being used as a research demonstration facility to lay the groundwork for a full scale demonstration plant that has recently been agreed to by a Chinese utility and Tsinghua University’s Institute of Nuclear Energy Technology.
A 110 MW(e) pebble bed modular reactor (PBMR) was proposed by Eskom, the South African Electric Utility, and was shut down in the spring of 2010. The PRINCIPAL ENGINEER has been directly involved in the research and development of this type of reactor technologies since his days as a graduate student in plasma physics at the University of Alberta.
The nuclear energy plant which is being developing by AscenTrust, LLC. is a modular, 110 Megawatt-electric (Mwe), high temperature, pebble bed reactor, using helium gas as a coolant and conversion fluid and gas turbine technology. The fundamental concept of the reactor is that it takes advantage of the high temperature and high pressure properties of the Brayton Cycle, using helium as a coolant. Use of the Brayton cycle in the production of electricity permit theoretical thermal efficiencies close to 50%. The other striking innovation of the NTPBMR is its inherent ability to extract all of the fuel elements from the core in the event of a loss of coolant.
Due to the NTPBMR’s innovative design approaches, advanced technologies and passive safety features, the safety assessment and the licensing of this reactor technology will require specific consideration, and the current NRC, LWR-based safety requirements will require special interpretation or adaptation.
The PRINCIPAL ENGINEER has set into motion a program to create a comprehensive set of Nuclear Safety Guidelines, and has even completed some of the documents. These Safety Guidelines have been mainly developed from the safety standards of water reactors, and their applicability to the NTPBMR is not always straightforward. For example, in NTPBMR designs, the fundamental safety functions are achieved with extensive use of passive and/or inherent features starting with the fuel element and the fuel delivery system to the reactor vessel. The implementation of defense in depth for the NTPBMR is quite different from that of water reactors. These differences will have significant impacts on the licensing approach for plant design, construction and operation.
Today’s operating nuclear plants were largely designed following a defense in depth strategy. According to INSAG-10, “Defense in depth consists of a hierarchical deployment of different levels of equipment and procedures in order to maintain the effectiveness of physical barriers placed between radioactive materials and workers, the public or the environment, in normal operation, anticipated operational occurrences and, for some barriers, in accidents at the plant. Defense in depth is implemented through design and operation to provide a graded protection against a wide variety of transients, including incidents and accidents, equipment failures and human errors within the plant and events initiated outside the plant”. This safety approach is reflected in the development of the NTPBMR Safety Standards for the design of the components of the nuclear power plants.
To provide guidance in licensing and safety assessments of NTPBMR’s, there is a need to develop an applicable set of safety requirements derived from the generally accepted principles of nuclear safety. The PRINCIPAL ENGINEER has recently developed a methodology for screening the defense in depth of nuclear power plants starting from the basic safety principles as proposed in INSAG-12. This methodology is used here to develop safety guidelines for the NTPBMR design and operation.
The objective of this document is to propose a technical basis and methodology, based on principles of defense in depth, for conducting design safety guidelines and in the long term generating design safety requirements for Pebble Bed Modular Reactor. The NTPBMR is used as an example to illustrate this process. For this purpose, the document provides an overview of the safety related features of current NTPBMR technology, examines how the defense in depth principle can be implemented/adopted by the NTPBMR design, and how the NTPBMR design could satisfy the three fundamental safety objectives:
• General nuclear safety
• Radiation protection
• Technical safety.
A discussion of these objectives and principles in Section 3 provides a framework for development of future publications related to the NTPBMR safety case.
This document focuses on the NTPBMR, as defined in Section 2. The NTPBMR incorporates some unique features. In particular the coated fuel particles, without metallic cladding, have the potential to retain radio-nuclides at temperatures well above their normal operating conditions, including the full range of design basis accident conditions. The helium coolant is an inert gas having no possibility of chemical interaction with other materials and no significant reactivity effects. For the Pebble Bed Modular Reactor, the decay heat is removed by thermal conduction, convection and radiation. The most interesting new feature is the ability to extract the fuel elements from the reactor vessel. As the fuel elements are being extracted from the reactor core they are being replaced by graphite spheres of the same dimension as the fuel element. This will cause a significant increase in the thermal mass of the core and ensure that there will be no internal rise in temperature of the core after the extraction of the fuel elements. The design uses simple and reliable passive means that ensure fuel temperatures are maintained within allowable limits even without reliance on the presence of the primary system coolant.
To apply the defense in depth screening approach, this document considered the three fundamental safety functions (control of reactivity, core heat removal, removal of heat from the extracted fuel elements and confinement of radioactive materials), and the challenges to the performance of these functions. Provisions identified are mainly based on design features of current PBMR and GT-MHR concepts, and are identified to illustrate the process for assessing the NTPBMR concept. This document does not consider challenges to the safety functions during various shutdown modes, or fuel storage and radioactive waste issues. A complete analysis, however, should also investigate all plant
Section 2 of this document presents a discussion of specific safety characteristics, particularly inherent safety features that form an integral part of the safety case. This discussion serves to define the family of concepts referred to in this document as the NTPBMR.
In Section 3, current general nuclear plant safety principles are addressed. Safety objectives, concepts and principles are described as a framework for design and operation of both current and future reactors. The structure of the existing nuclear safety standards is briefly described, identifying the role of the design requirements to ensure safety, and noting the logic underlying their development.
Section 4 introduces a method to prepare design safety guidelines for the NTPBMR, starting from the current requirements (mostly developed for light water reactors, LWRs), adopting a top-down approach applicable to the NTPBMR.
Section 5 presents a “critical review” of the reference requirements, analyzing the defense in depth implemented for the NTPBMR. For each level of defense in depth and for each fundamental safety function, the section illustrates the acceptance criteria for a successful achievement of the safety functions. The challenges to this successful behavior are identified as well as the mechanisms that originate the challenges. Finally the identification of the provisions to cope with these mechanisms create the basis for the definition of the design requirements.
Characteristics of reactor designs considered may be such that established LWR requirements are unnecessary, ineffective or even counterproductive. This requires an analysis of the specific design characteristics and safety features of the family of reactor designs and a full understanding of the role played by these features in achieving a safe design.
Section 6 summarizes the conclusions from the systematic investigation of the defense in depth of the NTPBMR, hopefully contributing to the future work of preparing design requirements for this family of future reactors. The second part of this sections provides a comparison of safety characteristics of LWRs and the NTPBMR.
SECTION TWO: NTPBMR DESIGN CHARACTERISTICS
2.1. NTPBMR DESIGN SAFETY CONCEPT
The NTPBMR’s fundamental safety objectives, requirements and design guidelines are based on the specific design characteristics and inherent safety features noted below:
• High quality ceramic coated-particle fuel of proven design, which adequately retains its ability to contain radioactive fission products over the full range of operating and accident conditions.
• The fuel micro-spheres are imbedded in a graphite matrix and coated with another layer of high quality ceramic coating to form what is essentially the fuel pebble.
• Post Loss of Coolant Accident (LOCA) shutdown passive removal of fuel pebbles from the core region of the reactor into a bath of Borated water, limiting maximum internal core temperatures in the LOCA event to values consistent with coated fuel particle and structural design limits.
• The Post Loss of Coolant Accident (LOCA) shutdown removal of fuel pebbles is done concurrently with the replacement of the fuel elements with graphite balls of the same diameter as the fuel element being extracted thereby lowering the maximum internal core temperatures in the LOCA event to values consistent with coated fuel particle and structural design limits.
• Post Loss of Coolant Accident (LOCA) shutdown passive injection of CO2 in the core will completely negate the possibility of air ingress into the reactor core.
• Post shutdown decay heat removal achievable through conduction, natural convection and radiation heat transfer, limiting maximum temperatures to values consistent with coated fuel particle and structural design limits.
• A single-phase inert coolant (helium), with no heat transfer limits that would be associated with phase change.
• Combination of low core power density, large reactor core and internals heat capacity, high core thermal conductivity and large fuel thermal margins, resulting in very long times (days) for evolution of response to loss of normal shutdown functions without protective actions.
• Fuel temperature margins and negative temperature-reactivity coefficients sufficient to accommodate any foreseeable reactivity insertions during startup and power operation without damage to the fuel
If successfully developed, the defining safety characteristic of the NTPBMR will be that its primary defense against serious accidents is achieved through its inherent design features. Active safety systems and prompt operator actions will increase the margins of safety since they are not required to prevent significant fuel failure and fission product release. The plant is designed such that its inherent features provide adequate protection despite operational errors or equipment failure. A primary design characteristic is the limitation of rated thermal power to a small fraction (on the order of 6 to 20%) of typical power levels for the large water reactors upon which the existing safety requirements are based. This is necessary to provide for removal of post shutdown decay heat using only passive means.
2.2. COATED FUEL PARTICLE
NTPBMR fuel is fully ceramic, and is therefore able to withstand much higher temperatures than can fuel elements with metallic cladding. The design of today’s coated fuel particle (CFP) has evolved empirically over several decades from a single layer of anisotropic carbon, to BISO (buffered isotropic pyrolytic carbon) to the current TRISO (triple isotropic layers) design. TRISO CFPs are small, typically ~1 mm diameter. In the TRISO design, the fuel kernel (typically Uranium-dioxide), is surrounded by a porous buffer layer of carbon to absorb fission gasses. Next there is an inner pyrolytic carbon (IPyC) coating; a silicon carbide (SiC) layer and then an outer pyrolytic carbon (OPyC) coating. Variations in CFP design are primarily in fuel type, kernel size, buffer and coating thickness and microstructure, and in methods for fabrication and quality control (QC) screening.
Since the CFP barriers form the primary line of defense against fission-product release, good performance is essential to the success of the NTPBMR design. For the most part, CFP designs have been arrived at empirically. A comprehensive analytical fuel performance model — accurately relating its (statistical) resistance to failure — has not been successfully developed due to the complexity of treating the combined effects of coating microstructure variations, variations in location and characteristics of microscopic imperfections, fission product chemical interactions along grain boundaries, fission gas pressure build-up, long term temperature and irradiation effects, etc. However, the empirical basis for CFP performance, a product of decades of development in many countries, is extensive.
An IAEA Co-ordinated Research Project (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behavior was conducted from 1992 to 1996, with participants from China, France, Germany, Japan, the Russian Federation, the United Kingdom and the USA. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for gas-cooled reactor (GCR) fuel performance and fission product behavior, and to verify and validate methodologies for the prediction of fuel performance and fission product transport.
CFP loss-of-function implies inability to retain fission products. Loss-of-function can range from long term diffusion of specific fission products (e.g. ceasium) through the coating layers, to sequential or simultaneous coating layer structural failure. There are many factors affecting fission product retention capability of any given CFP, including as-manufactured dimensions, coating layer microstructure, and chemical impurities; irradiation flux and temperature history, and chemical attack. In normal operation a particular concern for gas turbine (GT) designs is the diffusion of Ag-110m through the intact SiC layer at high operating temperatures. Silver deposition on turbine blades (and elsewhere) could lead to significant personnel exposure during maintenance. In accident conditions, CFP time/temperature history during the event tends to dominate the fission product release rate, particularly with regard to diffusion releases. Chemical attack from within (such as Palladium attack on SiC).
Diffusive release of several fission product species appears to begin at about 1600°C, although heating tests of irradiated CFP show very little release in the 1600°C area even for relatively long periods (typical of times at or near the peak in long-term depressurization accidents). Release rates increase markedly for time-dependent exposures in the 1700–2000°C range, and SiC degradation by chemical decomposition begins at approximately 2100°C; hence 1600°C is typically chosen as a conservative limit on peak fuel temperature under accident conditions. It should be noted also that predictions of peak fuel temperatures vs. time analyses often neglect to mention that a relatively small portion of the core fuel is at or near the peak (3-D time-temperature percent-fuel failure models account for this effect in core release predictions).
2.3. HELIUM AS PRIMARY COOLANT:
Helium gas pressurized to several MPa is employed as the primary system coolant. Helium is a single phase noble gas with no heat transfer limits associated with phase change. The absence of heat transfer limits (e.g. departure from nucleate boiling — DNB or critical heat flux — CHF) in addition to the core’s large thermal inertia will eliminate any safety related need to monitor short term variations in core power and temperature distributions. For the same reason, large local temperature increases during anticipated operational occurrences are less likely to occur. This will offer major operational benefits such as elimination or simplification of safety related monitoring and protection systems, and related surveillance and in-service inspection requirements.
In addition, due to the inert characteristics of helium, no significant chemical attack on fuel and other components would be expected if the contamination levels are kept low. Also, helium has no significant reactivity effects, and a relatively low amount of waste is generated due to activation and/or transmutation of the coolant impurities and corrosion products.
On the other hand, it is relatively easy for helium gas to leak from the primary circuit, especially at the elevated temperatures and pressures (although helium leakage does not cause any important safety issues). Thus for operational purposes, careful consideration is required for the design, fabrication, inspection and maintenance of the primary circuit. A monitoring system to detect the leakage should be able to identify leakage locations.
Helium will not condense if contained in a structure at normal temperatures following depressurization. Thus the pressure would reduce somewhat in accordance with the ideal gas law due to cooling, but would remain relatively high until the helium leaks out of the structure. In contrast, steam released from a water cooled system will condense on structural materials and components, resulting in a relatively rapid decrease in pressure. This characteristic substantially reduces the effectiveness of a conventional containment structure for a helium cooled system relative to a water cooled system. By retaining the helium following a depressurization, the gas leaking from the containment (typically specified as ≤1%/day for existing reactors) can serve as a transport mechanism for radionuclides which would be released from the fuel during a long term heat-up. Thus in many important scenarios a conventional containment would result in a higher offsite dose than a filtered vented confinement design.
2.4. DECAY HEAT REMOVAL VIA PASSIVE MEANS:
The NTPBMR designs relies on a passive ultimate heat sink system for removal of decay heat in the case of failure or unavailability of all active core cooling mechanisms. Under these conditions, core heat removal is accomplished via heat transfer from the core to the insulated reactor pressure vessel via conduction, radiation and (if coolant is present) convection, and from the vessel to the reactor cavity by radiation and convection. A reactor cavity cooling system (RCCS) is necessary to prevent overheating of the reactor cavity concrete during normal operation and to remove core decay heat under accident conditions.
The RCCS may not be necessary to prevent overheating of the fuel during accident conditions, as its unavailability would only cause a slight increase in peak fuel temperature. However, it may be necessary to prevent long term overheating of the reactor vessel and possible damage to or failure of reactor cavity structural elements and reactor supports. In typical designs the RCCS is fully operational during normal reactor operation, and there are no mechanical actions needed for it to function during a loss-of-forced-convection (LOFC) event. However, the operational mode may be different (e.g. transition from forced convection to natural convection RCCS cooling flow). Because of the multiple objectives and wide range of operational conditions, along with its necessarily massive size, the RCCS design and fabrication is challenging as well as crucial. In several instances, the performance of RCCS designs have been found to be difficult to predict with regard to local temperature distributions in the reactor cavity.
The heat load distributions for depressurized and pressurized LOFC accidents are quite different and may affect RCCS design requirements. For the depressurized case, the peak core temperatures tend to be near the level of the core beltline, while for the pressurized case, peak temperatures and heat loads are near the upper part of the vessel due to convection heating effects.
Additionally, accident analyses of some loss-of-cooling events for some designs have shown that a total functional failure of the RCCS has remarkably little impact on predicted peak fuel temperatures. However, variations among PBMR designs may significantly affect the functional requirements of the RCCS. For example, analyses have shown that for the higher power designs (~600 MW(t), RCCS operation is required during these accidents to protect the reactor pressure vessel from damage, while its failure does not necessarily lead to vessel damage for the lower power designs (~250 MW(t). Over the past two decades there has been a wide range of experimental and analytical work in this area in support of several PBMR designs. CRP on Heat Transport and Afterheat Removal for Gas Cooled Reactors under Accident Conditions was conducted from 1992 to 1997, with participants from China, France, Germany, Japan, Netherlands, the Russian Federation, and the United States of America.
The objective of this CRP was to establish sufficient experimental data at realistic conditions, and validated analytical tools to confirm the predicted safe thermal response of the NTPBMR during accidents. The scope included experimental and analytical investigations of heat transport by natural convection, conduction, and thermal radiation within the core and reactor vessel, and afterheat removal from the reactor vessel. Code-to-code and code-to-experiment benchmarks will be performed for verification and validation of the analytical methods which are being used.
2.5. LARGE THERMAL INERTIA, LOW POWER DENSITY, LARGE TEMPERATURE MARGINS
The combination of the NTPBMR core’s large thermal inertia (high heat capacity and low power density) typically results in long, slow core heat-up (and cool-down) transients for loss-of-forced-convection and loss of coolant pressure events. These attributes, coupled with the core’s high effective thermal conductivity attributes, tend to delay the occurrence of peak values of fuel temperatures for days, when the magnitude of the afterheat is considerably reduced. Very long response time also allows considerable opportunity for operational corrective measures to be taken.
The thermal response, in combination with the time-at-temperature effect on fuel fission product retention and the helium characteristics noted earlier, fundamentally alters the effectiveness of strategies for fission product containment. For example, in a depressurization accident, the predicted small fission product release from the fuel occurs long after the depressurization is completed, even for relatively small leaks. At this time, there would be no driving force to transport the fission products. In fact, once the maximum temperature is reached and the system begins to cool, the net flow is inward. However, if the released gas is contained, with a small (e.g. 1%/day) leakage rate, the leakage flow and slowly decreasing pressure would provide a mechanism for fission product transport. Thus attempting to contain the leaking helium can result in a higher fission product release rate for some of the most limiting events.
For annular core designs, the peak fuel temperatures in the depressurized accident scenarios (for a given total core power and vessel size) are reduced relative to those for a cylindrical active core. Increases in the core graphite conductivity, which can vary widely with irradiation and irradiation-temperature history, can also result in reduced peak fuel temperatures as the core graphite anneals, effectively increasing conductivity with increasing temperature. Thermal radiation effects also tend to become the dominating heat transfer mechanism for the pebble cores at the very-high (accident-range) temperatures.
2.6. TEMPERATURE MARGINS AND NEGATIVE TEMPERATURE-REACTIVITY COEFFICIENT
A negative temperature-reactivity coefficient can be attained in the NTPBMR for the entire fuel cycle and over the full temperature range of concern, as seen in most of the other types of reactors. In combination with the characteristics of large margin between fuel operation and fuel damage temperatures, and relatively low excess reactivity, as discussed below, power control and reactor shutdown can be ensured naturally. These characteristics significantly reduce the safety significance of the reactivity control and reactor shutdown systems. In the pebble bed reactor, the reactor core can operate with low excess reactivity by adjusting the number of fuel balls introduced during operation. Protection and management of abnormal reactivity insertion conditions could be provided by inherent features, simplifying the design of active/passive protection or mitigation systems to assure safe shutdowns.
2.7. FEATURES COMMON TO NTPBMRs AND OTHER FUTURE REACTORS
A. Simplification and use of passive systems: NTPBMRs make extensive use of passive characteristics that offer the opportunity to eliminate or simplify active systems that rely on a large number of safety grade support systems by applying the advantages of simple gravity driven or thermal gradient driven safety systems. The challenge is to demonstrate the capability and the reliability of these passive systems, in particular for the long time accident response.
B. Standardization, prefabrication and modularity: The standardization, prefabrication and modularity of the facilities that will likely be part of the design, construction and operation of NTPBMR with evident benefits on the economics of a single unit, will also lead to a simplification of the licensing through a certification procedure, and reduction of the construction time and licensing costs.
C. Applicability of PSA and risk-informed decision making: Because of the extensive use of passive components, the safety of these reactors is primarily determined by initiating events of very low probability (e.g. structural failures due to extremely rare external events). The consequences of these events are determined by the direct phenomenological response of the plant to these events, rather than by a sequence of failures of systems, which individually have higher probabilities and which can be analyzed and modelled with much less uncertainty. This aspect will pose significant challenges for the development and application of PSA methodologies to address these concepts.
SECTION THREE: GENERAL SAFETY ASPECTS OF NUCLEAR POWER PLANTS
3.1. SAFETY OBJECTIVES
The Safety of Nuclear Power Plants: Design, sets out basic objectives, concepts and principles for ensuring safety of nuclear installations in which the stored energy or the energy developed in certain situations could potentially result in the release of radioactive material from its designated location with the consequent risk of radiation exposure of people. The principles are derived from the following three fundamental safety objectives:
1. General Nuclear Safety Objective: To protect individuals, society and the environment from harm by establishing and maintaining in nuclear installations effective defenses against radiological hazards. This general nuclear safety objective is supported by two complementary safety objectives dealing with radiation protection and technical aspects. They are interdependent: the technical aspects in conjunction with administrative and procedural measures ensure defense against hazards due to ionizing radiation.
2. Radiation Protection Objective: To ensure that in all operational states radiation exposure within the installation or due to any planned release of radioactive material from the installation is kept below prescribed limits and as low as reasonably achievable, and to ensure mitigation of the radiological consequences of any accidents
3. Technical Safety Objective: To take all reasonably practicable measures to prevent accidents in nuclear installations and to mitigate their consequences should they occur; to ensure with a high level of confidence that, for all possible accidents taken into account in the design of the installation, including those of very low probability, any radiological consequences would be minor and below prescribed limits; and to ensure that the likelihood of accidents with serious radiological consequences is extremely low.
Safety objectives require that nuclear installations are designed and operated so as to keep all sources of radiation exposure under strict technical and administrative control. However, the radiation protection objective does not preclude limited exposure of people or the release of legally authorized quantities of radioactive materials to the environment from installations in operational states. Such exposures and releases, however, must be strictly controlled and must be in compliance with operational limits and radiation protection standards.
In order to achieve these three safety objectives in the design of a nuclear power plant, comprehensive safety analyses are carried out to identify all sources of exposure and to evaluate radiation doses that could be received by the public and by workers at the installation, as well as potential effects of radiation on the environment. The safety analysis examines:
1. All planned normal operational modes of the plant
2. Plant performance in anticipated operational occurrences;
3. Design basis accidents
4. Selected severe accidents.
The design for safety of a nuclear power plant applies the principle that plant states that could result in high radiation doses or radionuclide releases are of very low probability of occurrence, and plant states with significant probability of occurrence have only minor or no potential radiological consequences. An essential objective is that the need for external intervention measures may be limited or even eliminated in technical terms, although such measures may still be required by national authorities.
3.2. THE DEFENSE IN DEPTH STRATEGY
The safety objectives will be achieved through the application of the defense in depth strategy. The strategy for defense in depth is twofold: first, to prevent accidents and, second, if prevention fails, to limit their potential consequences and prevent any evolution to more serious conditions. Accident prevention is the first priority. The rationale for the priority is that provisions to prevent deviations of the plant state from well-known operating conditions are generally more effective and more predictable than measures aimed at mitigation of such departure, because the plant’s performance generally deteriorates when the status of the plant or a component departs from normal conditions. Thus preventing the degradation of plant status and performance generally will provide the most effective protection of the public and the environment as well as the protection of the investment.
Should preventive measures fail, however, control, management and mitigatory measures, in particular the use of a well-designed confinement function, can provide the necessary additional protection of the public and the environment. The concept of defense in depth, as applied to all safety activities, whether organizational, behavioral or design related, ensures that they are subject to functionally redundant provisions, so that if a failure were to occur, it would be detected and compensated for or corrected by appropriate measures. Application of the concept of defense in depth in the design of a plant provides a series of levels of defense (inherent features, equipment and procedures) aimed at preventing accidents and ensuring appropriate protection in the event that prevention fails. This strategy has been proven to be effective in compensating for human and equipment failures, both potential and actual. There is no unique way to implement defense in depth (i.e. no unique technical solution to meet the safety objectives), since there are different designs, different safety requirements in different countries, different technical solutions and varying management or cultural approaches. Nevertheless, the strategy represents the best general framework to achieve safety for any type of nuclear power plants.
Generally, several successive physical barriers for the confinement of radioactive material are put in place. Their specific design may vary depending on the activity of the material and on the possible deviations from normal operation that could result in the failure of some barriers. So, the number and type of barriers confining the fission products is dependent on the adopted reactor technology.
Defense in depth is generally structured in five levels. Should one level fail, the subsequent level comes into play. The list below summarizes the objectives of each one of the five levels and the correspondent primary means of achieving them. The general objective of defense in depth is to ensure that a failure, whether equipment failure or human failure, at one level of defense, and even combinations of failures at more than one level of defense, would not propagate to defeat defense in depth at subsequent levels. The independence of different levels of defense, i.e. the independence of the features implemented to fulfill the requested functions at different levels, is a key element in meeting this objective.
LEVELS OF DEFENSE IN DEPTH (FROM INSAG-10)
• Level 1: Prevention of abnormal operation and failures. This essentially means Conservative design and high quality in construction and operation.
• Level 2: Control of abnormal operation and detection of failures. This essentially means control, limiting and protection systems and other surveillance features.
• Level 3: Control of accidents within the design basis. This essentially means engineered safety features and accident procedures.
• Level 4: Control of severe plant conditions including prevention of accident progression and mitigation of the consequences of severe accidents. This essentially means complementary measures and accident management.
• Level 5: Mitigation of radiological consequences of significant releases of radioactive materials. This essentially means off-site emergency response.
Note: For existing plants, the term ‘severe accidents” is widely associated with significant melting of the core and large releases of radio-nuclides from the reactor vessel. Because of the characteristics and features of the NTPBMR discussed in Section 2, and in particular the low core power density, high temperature capability of the coated fuel particles and the system’s ability to extract the fuel elements from the core, no scenarios involving extensive melting of the core are apparent, even for very low probabilities/highly hypothetical events. Thus in the case of the NTPBMR, the term ‘severe accident’ is taken to mean events which could challenge the structural integrity of the core and thus the ability to predict the course of the event, e.g. sustained (days) air ingress through large openings in the primary system and the confinement building. However, some action to manage these situations would be advisable to maintain the plant in a state that can be analyzed. While such conditions could serve as a basis for considerations associated with Level 4 of defense in depth, it is important to point out that these extreme conditions will not necessarily involve large releases from the fuel, since existing data show effective radionuclide retention at elevated temperatures when the fuel has burned back to the silicon carbide layer of the coated particles and remains in a high temperature air environment for days.
3.3. THE FUNDAMENTAL SAFETY FUNCTIONS
The objective of the safety approach is to provide adequate means:
• To maintain the plant in a normal operational state;
• To ensure the proper short term response immediately following a postulated initiating event (PIE);
• To facilitate the management of the plant in and following any design basis accident, and following any plant states beyond the design basis that may occur (i.e. the “severe plant conditions”).
To ensure safety (i.e. to meet allowable radiological consequences during all foreseeable plant conditions), the following fundamental safety functions shall be performed in operational states, in and following a design basis accident and in and after the occurrence of severe plant conditions:
• Control of the reactivity
• Removal of heat from the core;
• Confinement of radioactive materials and control of operational discharges, as well as limitation of accidental releases.
The possible challenges to the safety functions are dealt with by the provisions (inherent characteristics, safety margins, systems, procedures) of a given level of defense. Combinations of one or more provisions to cope with challenges to levels of defense are often called lines of defense (LOD). The way the fundamental safety functions are achieved and the specific LOD used, are obviously dependent on the specific design.
All mechanisms that can challenge the successful achievement of the safety functions are identified for each level of defense. These mechanisms are used to determine the set of initiating events that encompass the possible initiations of sequences. According to the philosophy of defense in depth, if the evolution of a sequence is not controlled by the provisions of a level of defense it will be by the subsequent level that comes into play (LOD functional redundancy).
The objective of Defense in Depth is always to maintain the plant in a state where the fundamental safety functions (confinement of radioactive products, control of reactivity and heat removal) are successfully fulfilled. Success criteria are defined for each level of defense in depth and for the moment they are expressed only in deterministic terms. As the objective of the first level of protection is the prevention of abnormal operation and system failures, if it fails, an initiating event comes into play and a sequence of events is potentially initiated. Then the second level of protection will detect the failures or control the abnormal operation. Should the second level fail, the third level ensures that the safety functions are further performed by activating specific LODs (safety systems and other safety features). Should the third level fail, the fourth level limits accident progression through accident management, so as to prevent or mitigate severe accident conditions with external releases of radioactive materials. The last level (fifth level of protection) is the mitigation of the radiological consequences of significant external releases through the off-site emergency response. Some challenges/mechanisms may compromise the effectiveness of the considered level of defense by affecting either the performance of the safety function directly or the reliability of a safety provision. The effectiveness of a level of defense is determined by the ability of the provisions to cope with mechanisms which challenge the performance of safety functions. The probability associated with challenges/mechanisms, the reliability of the demanded safety provisions and the associated potential radiological consequences will define the risk for the considered accident sequence.
3.4. THE CONCEPT OF LINES OF DEFENSE
To evaluate or compare the implementation of defense in depth by different reactor technologies, it is suggested to adopt a common approach that needs to have the following features:
• The safety objectives should be the same in terms of doses respectively to the operators, the public and the environment (i.e. radiological consequences) for all plant conditions at a given level of defense;
• The safety assessment method should use analogous and comparable approaches based on the integral adoption of defense in depth (all the levels should be considered);
• The approach should be able to integrate the unique characteristics of each type of reactor, with the number and the quality of the required “defenses” being a function of the potential internal and external hazards and consequences of failures.
To implement this, it is useful to introduce the concept of lines of defense as any inherent characteristic, equipment or system implemented into the safety related plant architecture, as well as any safety relevant operational procedure, that are necessary to fulfil the safety functions. The required number and strength of these lines of defense depend on the reactor type, i.e. the implemented LODs shall fulfil the missions requested to prevent abnormal situations or return the plant to a controlled or safe shutdown condition and maintain it in a safe state after a postulated initiating event (PIE). Their design shall take into account simultaneously the needs for performance (to meet the safety criteria), and the safety objectives as well as the recommendations concerning, for example, reliability, redundancy, diversity, in-service inspection requirements, etc.
In this logic, the physical barriers normally considered in LWRs (fuel, cladding, primary circuit and containment) are provisions to confine fission products. Their contribution to safety has to be assessed for each specific concept of reactor and considered in the general safety architecture of the plant. As lines of defense can rely simultaneously on both active and passive systems as well as on inherent features, the safety assessment approach should consider their correspondent reliabilities to correctly take into account all the potential of the safety related architecture. The LODs can be classified into categories according to their reliability. The number and category of LODs can be used as a tool to assess the adequacy of the implementation of defense in depth.
3.5. CURRENT SAFETY APPROACH
Operating nuclear power plants are largely designed following a safety architecture dictated by the implementation of the strategy of defense in depth (physical barriers and levels of defense) as illustrated in Section 3.2. In the majority of the plants of the current generation the application of defense in depth is mainly based on deterministic considerations. This means that the plant is deterministically designed against a set of normal and postulated accident situations according to well established design criteria in order to meet the radiological targets. The adequacy of the defense in depth is established by the number of barriers and number and quality of systems in each level of defense.
The current design approach has been shown to be a sound foundation for the safety and protection of public health, in particular because of its broad scope of accident sequence considerations, and because of its many conservative assumptions which have the effect of introducing highly conservative margins into the design that, in reality, give the plant the capability of dealing with a large variety of sequences, in some cases well beyond those included in the design basis. The deterministic approach is complemented by probabilistic evaluations with the main purpose of verifying that the design is well balanced and there are no weak areas or systems that could allow for the possibility of high risk sequences. Probabilistic safety assessment is recognized as a very efficient tool for identifying those sequences and plant vulnerabilities that require specific additional preventive or mitigative design features.
3.6. THE ASCENTRUST SAFETY OBJECTIVES
The PRINCIPAL ENGINEER of AscenTrust has given himself the task of establishing standards of safety for protection against ionizing radiation for its stakeholders and strategic partners. The design related documents will be issued in the Safety Objectives, covering nuclear and radiation safety.
Safety Fundamentals: present basic objectives, concepts and principles of safety and protection in the development and application of the NTPBMR technologies to the production of electricity through the use of nuclear power.
Safety Requirements: establish the requirements that must be met to ensure safety. These requirements, which are expressed as ‘shall’ statements, are governed by the objectives and principles presented in the Safety Fundamentals.
Safety Guides: recommend actions, conditions or procedures for meeting safety objectives. Recommendations in Safety Guides are expressed as ‘should’ statements, with the implication that it is necessary to take the measures recommended or equivalent alternative measures to comply with the requirements.
3.7. DEVELOPMENT OF GENERAL SAFETY DESIGN OBJECTIVES
Design requirements play an important role in establishing the safety level of the installation and also have great impact on its cost and operating procedures. The general logical process to generate the safety requirements for a reactor plant design is briefly described below. The Safety requirements can be derived from a set of limited safety principles which directly descend from the three well established safety objectives. The safety objectives define the general targets that shall be achieved by a nuclear installation to protect the operators and the population. They are the same for all nuclear installations including nuclear reactors, and are independent of the kind or size of any given installation.
For nuclear reactors, the compliance with the safety objectives is achieved when the three fundamental safety functions Confinement of radioactive material, control of the reactivity and removal of the heat from the core are fulfilled for all the plant operational, accidental and post accidental conditions in accordance with radiological targets. To ensure that the safety objectives are met with sufficient confidence and the fundamental safety functions are adequately fulfilled, an effective defense in depth should be implemented. For measuring and assessing the adequacy of the defense in depth, success criteria (expressed in deterministic and probabilistic terms) need to be defined for each level of defense.
Defense in depth has been proved to be generally applicable and very effective in assuring safety in NPPs. It can be used as primary guidance for the preparation of safety requirements. It is reasonable to assume that this correspondence is maintained for all kind of reactors regardless of their size or specific safety features. The safety requirements can be obtained by developing, for each fundamental safety function, the corresponding provisions necessary to meet the established success criteria for each level of defense. The correct implementation of the strategy of the defense in depth (i.e. the adoption of an adequate safety architecture) ensures that the fundamental safety functions are reliably achieved and with sufficient margins to compensate for equipment failure and human errors. More demanding success criteria will result in a more effective defense in depth and in more demanding requirements for the provisions for each level of defense.
3.8. INTEGRATION OF DETERMINISTIC AND PROBABILISTIC CONSIDERATIONS IN THE SCHEME OF DEFENSE IN DEPTH
The generalized concept of defense in depth, as outlined in Section 3.2, needs to integrate both deterministic and probabilistic considerations (e.g. system reliability probabilistic targets, etc.) to provide metrics for assessing the adequacy of the means of each level of defense. The integration of deterministic and probabilistic approaches also provides a basis for additional requirements and to ensure a well-balanced design.
The actual level of safety is determined by the full set of detailed criteria and requirements (deterministic and probabilistic) with which the design complies. In other words, the level of safety depends on the way defense in depth is implemented in the design taking into account the implications of the specific features and technology. The approach provides general guidance on what is understood to be key engineering judgments about the performance requirements of the plant systems. However, the levels of defense by themselves do not provide the metrics by which to judge adequacy of the implementation of defense in depth. Risk informed approaches which combine deterministic and probabilistic techniques, can be useful tools to assess the contribution of each line of defense to safety with a resulting integrated safety assessment relative to public health and safety.
The approach that is recommended is the development of a probabilistic safety assessment model of all plant systems without any pre-conceived notion of what is safety related. This model can then be used to determine the importance to safety of systems, structures and components which can then lead to a determination of safety classification. This model can then also be used to assess the contribution of each level of defense to the ultimate safety of the plant as it relates to public health and safety. Should there be barriers or other provisions that need to be strengthened, the value of the improvement can be directly assessed.
A key factor in making safety adequacy assessments is the ability to tie the levels of defense concept to safety goals that are generally accepted for nuclear plants. This linkage provides the integration of safety with technology judgments of adequacy from a public health and safety point of view. The risk informed process can be used in plant design to optimize safety performance and to balance the lines of defense in an overall defense in depth strategy by the quantification possible through the use of probabilistic safety analysis. One of the key issues in deterministic and probabilistic analysis is how to deal with uncertainties. Traditional deterministic approaches rely on a balance of prevention and mitigation with large design margins and the ultimate final barrier being the ‘containment’ to cover any unknown phenomenon or event that goes beyond what is generally expected or understood. With the pebble bed reactor, the objective is to design the plant making extensive use of inherent safety features that do not rely on active systems to prevent plant conditions that could lead to fuel failure and fission products release. By employing the risk informed analysis, the contribution to safety of the design features and need for additional features can be assessed. To deal with uncertainties, especially in early deployment of the systems, sensitivity analysis of the performance of key systems can be used to provide a measure of the impact of the uncertainty and appropriate design decisions can be made.
SECTION FOUR: DESIGN SAFETY OBJECTIVES FOR THE NTPBMR
4.1. THE TOP-DOWN APPROACH
The proposed top-down approach consists of a systematic review of the existing requirements for nuclear power plants starting from the most general (applicable to all nuclear plants) and down to the most specific and more technology dependent.
The requirements for the NTPBMR are generated through a critical interpretation of the objectives, challenges to the objectives, mechanisms posing the challenges and corresponding provisions associated with each level of defense in depth and the full understanding of the safety features of the NTPBMR reactor. The safety requirements for nuclear power plants have reached the current status through a long development process which incorporated the results of the extensive operating experience and the experience gained from the errors of the past. The current safety requirements define the safety approach developed and refined over many years. Although they are mostly developed for water cooled reactors, it is reasonable to assume that they are a good starting point for the preparation of the design requirements for any type of reactors including non-water cooled reactors such as the NTPBMR. For this technology, which make extensive use of inherent safety features, it can be expected that the acceptance criteria of each level of defense could be met using less and simpler safety systems than those for large water reactors.
The mechanism for judging the applicability or adequacy of a requirement for existing NPPs to the NTPBMR should be based on the full understanding of its contribution to defense in depth. The ‘transfer function’ that establishes the requirements for a generic nuclear reactor plant from the requirements for existing plants, should not simply be interpreted as a filter to accept or not a requirement but as a mechanism to generate new requirements if they are necessary because of the features of the specific plant. For example, an inherent feature that fulfils a safety function in a very reliable way could allow for a relaxation of the requirements for a safety system or even to the possible elimination of the safety system that performs an equivalent function for water reactors. On the other hand, the designer should be aware that specific features or materials could possibly initiate events for which adequate preventive or mitigative measures could be necessary. This process will lead to the compilation of a consistent set of requirements organized in a hierarchical way with the general requirements at the top and the more specific at the bottom like those existing for current plants.
4.2. APPLICABILITY OF CURRENT DESIGN REQUIREMENTS TO THE NTPBMR
The current design requirements have been mainly developed for water reactors, and their applicability to the design of the NTPBMR is not always straightforward. In some cases, special interpretation may be necessary. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants (NPPs). They must be met for safe operation of an NPP, and for preventing or mitigating the consequences of events that could jeopardize safety.
The Design Control Document, the first essential document shall be used as a baseline reference to the safety case for the NTPBMR will be developed and attached to this document as a Reference. This document will include requirements for a comprehensive safety assessment to be carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states:
• Elaborates on the three safety objectives and the concepts like defense in depth which form the basis for deriving the safety requirements that must be met in the design of any NPP.
• Covers the requirements to be applied by the design organization in the management of the design process, and also the requirements for safety assessment, for quality assurance, and for the use of proven engineering practices and operational experience. These principal requirements should be applicable to any NPP design independent of the technology adopted.
• Provide the general technical requirements for defense in depth and radiation protection. They should be also independent of the adopted technology.
• Provides the requirements that are applicable to the process of the design itself. It covers safety classification, general design basis, design for reliability, provisions for in-service testing, maintenance and repair, equipment qualification, ageing, human factors, safety analysis and other considerations. Although the implementation of the requirements will conduct to technology dependent solutions (e.g. considered PIEs, in-service inspection solutions, etc.), the requirements are generically stated and, therefore, they are applicable to any type of reactors.
• Provide design requirements applicable to specific plant systems, such as: the reactor core and associated features, reactor coolant systems, containment systems, instrumentation and control, fuel handling and storage system. These are the most technology-dependent requirements and a deeper investigation should be conducted to determine to what extent they need adaptation or modification for NTPBMR designs.
4.3. THE OBJECTIVE-PROVISIONS TREE
The method of the objective-provisions tree, represents a preliminary attempt to systematically address the “critical review” of the implementation of the defense in depth. The logical framework of the objective-provisions method is graphically depicted in terms of a tree. At the top of this tree is the level of defense in depth of interest, followed by both the objectives to be achieved and the barriers or defenses to be protected.
For each sub-function, the challenges to its fulfilment can be identified. These challenges are general processes or situations that can prevent adequate performance of the safety functions (e.g. reactivity excursions that could damage the fuel before the shutdown). The challenges arise from a variety of mechanisms (or events) which also have to be identified. The identification of the mechanisms (or events) that can challenge the success of a safety function is an essential task in the development of the logical framework for inventorying the defense in depth capabilities of a nuclear power plant. Once the mechanisms are understood, it is possible to determine the provisions necessary to prevent and/or control these mechanisms. If the set of provisions of a Level N is not sufficient to overcome some mechanisms of a challenge to the safety function or some failures prevent the provisions to perform their function, then additional provisions will come into play to support safety functions to achieve acceptance criteria correspondent to the subsequent Level N+1.
SECTION FIVE: IMPLEMENTATION OF DEFENSE IN DEPTH FOR THE NTPBMR
In this section, general characteristics of the NTPBMR drawn from existing designs and potential provisions based around these characteristics are used to explore the implementation of defense-in-depth using the methods identified in this document. The considerations presented here are intended to illustrate application of the methods and are not intended to be requirements for the NTPBMR. However, they can be viewed as a first step in the development of the requirements.
5.1. GENERAL CONSIDERATIONS ON BARRIERS AND LEVELS OF DEFENSE IN DEPTH
The implementation of defense in depth (D.i.D.) for the NTPBMR differs from that for the traditional LWR strategy to achieve effective defense against radiological hazards. The safety of the NTPBMR relies strongly on inherent features, with the confinement of radionuclides being accomplished with minimal or no reliance on active systems or operator actions. Using the definition in INSAG-10, defense in depth consists of a hierarchical deployment of different levels of equipment and procedures (LOD) in order to maintain the effectiveness of physical barriers placed between radioactive materials and workers, the public and the environment in normal operation, anticipated operational occurrences and, for some barriers, in accidents at the plant. Defense in depth is implemented through design and operation to provide a graded approach to defense in a wide variety of transients, incidents and accidents, including equipment failures and human errors within the plant as well as events initiated outside the plant.
The public and the environment are protected primarily by means of these barriers, which may serve both operational and safety purposes or safety purposes only. The defense in depth concept applies to the protection of their integrity against internal and external events that may jeopardize it. Situations in which one or more barriers are breached (such as during shutdown) may require special attention. The description of the “barriers” that can be identified in the NTPBMR requires special attention because their importance to safety may vary relative to water reactors. A proposed definition of the barriers is as follows: the kernel (i.e. the fuel material), three particle coating layers, the matrix (i.e. the graphitic material around the particles), the fuel element (i.e. pebble/fuel assembly block), the primary circuit, the plant civil/structural/confinement works, and the filtering system(s). It should be noted that some of these barriers (e.g. pyrocarbon coatings, matrix) are not impervious to all fission products (e.g. cesium), even when intact, and the effectiveness of these barriers in confining radioactive material varies widely, and is dependent on operating (normal/accident) conditions. In the NTPBMR, the primary barrier is the silicon carbide layer of the coated fuel particle. There are other “barriers” that reduce the release of fission products into the environment. These other barriers as noted above are effective contributors to the defense in depth of the NTPBMR design to limit the release of radioactive materials into the environment and dose to the public.
Concerns have been expressed about the effectiveness of the coated fuel particle (CFP) in providing a containment function, since there are literally billions of them involved in the process. However, for the NTPBMR designs, the unique characteristics of the technology allow for important complementary considerations that can further enhance the strength and resilience of the robust safety case. The kernel and coating layers of the coated fuel particle (CFP) constitute successive barriers operating in parallel among the billions of particles comprising a typical NTPBMR core, with each particle containing an insignificant amount of fission products. This population of parallel barriers cannot act in a uniform way in any conceivable circumstance because of the following variations:
• Variations within a batch – The nature of the fuel kernel production and fluidized bed coating processes result in a statistical variation of kernel and coating properties such as kernel diameter and coating thickness within a given batch. Mean values and standard deviations in these properties are specified as a part of the fuel product acceptance criteria.
• Fabrication batches – the core at any given time will consist of hundreds of combinations of kernel, coating and fuel compact or sphere fabrication batches.
• Service conditions – the core at any given time will generally consist of a population of particles with a broad range of service conditions. Spatial variations in temperature and neutron flux, as well as variations in time of service, will produce a broad range of particle histories for the key parameters of temperature history, fluence and burnup.
• Event conditions – The extent of the challenge to the containment barriers of a given coated particle is determined by its service conditions as well as by the conditions experienced in a given event. The most important event condition is particle temperature, which will vary over a wide range in any event, with the population mean temperature far below the maximum temperature.
This diversity effectively addresses concerns about fuel that would perform well in normal operation and yet suddenly fail at lower-than-expected temperatures, with the possibility of a sudden onset of barrier failures in a large fraction of the particle population under accident conditions (‘weak fuel’).
Additional defense in depth considerations involve the preservation of the effectiveness of the barriers and options for dealing with the loss of barrier functions. The current safety practice allows for the loss of some barrier function within the design basis and selected severe accidents of a nuclear plant, with a requirement that at least one of the barriers should remain effective and contain the fission products to ensure compliance with the radiological targets. The safety design of the NTPBMR technology and its operation is consistent with this logic, as discussed further in this section. Measures relative to defense in depth are ranked in five levels of defense. The first four levels are oriented towards the protection of barriers and mitigation of releases; the last level relates to off-site emergency measures to protect the public in the event of a significant release. Even though implementation of the concept of defense in depth may differ from LWR to the NTPBMR and may to a certain degree depend on plant design, the main principles are common.
For a consistent implementation of the defense in depth concept, account needs to be taken of the risk represented by the amount and type of radioactive material present in the installation; the potential for its dispersion due to the physical and chemical nature of these products; and the possibility of nuclear, chemical or thermal reactions that could occur under normal or abnormal conditions, and the kinetics of such events. These complementary considerations that differentiate the technology of the NTPBMR from other kinds of reactors, in particular for the characteristics of the fuel, can be referred to as ‘defense in breadth’. installation; the potential for its dispersion due to the physical and chemical nature of these products; and the possibility of nuclear, chemical or thermal reactions that could occur under normal or abnormal conditions, and the kinetics of such events.
The method of objective-provisions trees is adopted here to systematically conduct the ‘critical review’ of the implementation of defense in depth for the NTPBMR designs. For each one of the first four levels of defense, three objective-provisions trees will be developed correspondent to the three fundamental safety functions. With respect to Level 5, reliance on off-site measures to mitigate consequences of severe accidents should be minimal due to the effectiveness of the previous levels of defense. As provisions of Level 5 of defense in depth do not normally involve design, they are outside of the scope of the present Document. The adopted strategy to implement defense in depth for the NTPBMR differs from the traditional LWR philosophy and gives higher priority both to the prevention of accidents through a significant plant architecture simplification that minimizes the number of failures with potential safety significance and to the management of abnormal situations through the implementation of robust LODs (e.g. TRISO particle, passive DHR, etc.). These aspects put strong emphasis on Level 1 and Level 3 of the defense in depth, and considerably enhance the robustness of the overall safety case.
5.2. APPLICATION OF LEVEL 1 DEFENSE IN DEPTH FOR THE NTPBMR
The objective for Level 1 is the prevention of deviations from normal operation, the prevention of failures, and to ensure that the safety systems would operate reliably if called upon at higher levels of defense. The essential means are the provision of the characteristics described in Section 2, conservative design, and high quality in construction and operation. A primary means for preventing accidents is to strive for such high quality in the design that deviations from the normal operation states are well within prescribed design limits. As for other kind of NPPs, a large number of deviations from normal operation can be avoided through adequate site selection which reduces the likelihood of externally initiating events, either natural or human-induced. Challenges to the safety functions due to unexpected mechanical loads should be compensated by a conservative structural design which takes into consideration loads originated by external events.
Prerequisites for safe operation are careful selection of materials and use of qualified fabrication processes and proven technology, together with extensive testing. In this aspect, the NTPBMR design will incorporate well known and proven structural materials, high purity graphite for core internals and high quality ceramic coated particle fuel of proven design, together with recent technological advances in areas like magnetic bearings, compact plate-fin heat exchanger and turbo-machine development. For the NTPBMR, a safety function of the vessel system is to ensure that the core geometry is maintained within acceptable limits under all normal and postulated abnormal conditions. This safety function is derived from two of the fundamental safety functions, named core heat removal and control of reactivity. During normal operation conditions, with insured core geometry, the heat removal is performed by a helium coolant system using reliable turbo-compressors. An adequate conservative design of core support and barrel structure provides support and alignment for the components that are housed within the reactor vessel. This will avoid insertions of reactivity by preventing changes in core geometry and will also ensure the ability of control rods to insert and safely shut down the reactor. Support and restraint structures are considered part of the pressure boundary system. The pressure boundary is designed to an international pressure vessel code or standard capable of ensuring that all of the functional, safety and reliability requirements can be met. Provisions can be made for the replacement of some or all of the reactor internals, depending on the degree of confidence in the component lifetimes and reliability. Inspections can be carried out on the ceramic and metallic parts. These preventive surveillance and maintenance measures are also considered as part of the safety provisions at defense in depth Level 1. For reactivity induced events, the absence of a steam loop, the higher pressure of helium circuits relative to water cooling systems, and specific design solutions to exclude the presence of water sources, completely negates the possibility of water ingress into the core. Furthermore, unexpected reactivity insertion due to malfunction of the Reactivity Control System is also minimized by seismically designed units which operate under a fail-safe mode. The possibility of operator induced failures is also reduced by design thermal margins, slow thermal response, and other inherent features which minimize or simplify demands for manual intervention.
The NTPBMR safety philosophy is based on control of releases primarily by the retention of radionuclides within the coated fuel particle rather than reliance on secondary barriers (such as the primary coolant boundary or the reactor building). Thus, ensuring that the safety criteria are met is the same as ensuring that the retention capability of the coated fuel particles (CFP) is not compromised. There is a considerable design margin between normal operation service conditions and fuel failure temperature. The importance of the safety function of the pressure boundary system to contain the helium coolant by maintaining vessel integrity is reduced by the ability of the designs to remove decay heat without reliance on the presence of the helium coolant. To ensure fuel integrity under normal operation conditions, design requirements limit chemical and other physical attack on the fuel. The chemically inert helium coolant also minimizes corrosion and eliminates complications associated with internal cladding of the vessel walls. The function of the helium purification system is to remove chemical and particulate contaminants from the primary coolant in order to provide the necessary degree of helium purity during normal operation as well as removing small amounts of fission or activation products present in the coolant in normal operation. Low induced activity of helium is another factor in the low level of radiological consequences expected from leakage during normal operation. The use of magnetic bearings in the turbomachinery can eliminate coolant contamination by lubricating oil. The radiological design objective is that for all pathways any dose received by the operators and the public and releases to the environment in normal operations will not only meet regulatory limits and constraints, but will also be as low as reasonably achievable (ALARA principle). The NTPBMR design will minimize the generation of radioactive waste throughout its lifecycle (including decommissioning) and include appropriate processing, conditioning handling and storage systems.
Another typical measure for Level 1 is the provision, as a design attribute, for adequate time for operators and the system to respond to normal events. In an overall sense, the robustness of the NTPBMR designs is achieved through combination of single phase coolant, low power density, high core heat capacity, and large temperature margins between normal operation and fuel failure temperatures. These Level 1 characteristics provide a more stable operation, and may reduce the requirements for Level 2, which dictate the accuracy, response time, and reliability requirements of the control and protection systems. Additional typical operating measures corresponding to Level 1, for the safe operation of both LWR and the NTPBMR, are:
• Comprehensive training of appropriately selected operating personnel whose behavior is consistent with a sound safety culture;
• Safety monitoring system connected real time with the prototype will ensure that all deviations from normal operating conditions will be immediately noticed.
• Adequate operating instructions and reliable monitoring of plant status and operating conditions
• Comprehensive preventive maintenance prioritized in accordance with the safety significance and reliability requirements of systems.
The objective-provisions tree for the safety function of control of reactivity where, for the objective of Level 1 of defense in depth, the acceptance criteria are stated as:
• To avoid insertion of reactivity which demands countermeasures outside the normal control range;
• To ensure the ability to safely shutdown the reactor during normal operation, anticipated operational occurrences and design basis accidents.
The objective-provisions tree for the safety function of core heat removal, with the Level 1 corresponding to acceptance criteria being:
• To transfer the power generated in the core to the balance of plant (BOP), respecting allowed temperature ranges on fuel and structures during normal operation;
• To ensure the ability to safely remove the decay heat during normal operation, anticipated operational occurrences and design basis accidents.
The objective-provisions tree for the safety function of confinement of radioactive material, where the acceptance criteria for Level 1 are:
• Concentration of radionuclides (including fission products) below the limits established for normal operating conditions in the reactor coolant system and inside the reactor building;
• Ensure the ability of maintaining barriers for confining radioactive materials for normal operation, anticipated operational occurences and design basis accidents.
The provisions identified for Level 1 address both those which are necessary to support normal operation and those necessary to assure the capability to perform the key safety functions during anticipated operational occurrences and design basis accidents.
5.3. APPLICATION OF LEVEL 2 DEFENSE IN DEPTH FOR THE NTPBMR
The objective for Level 2 of Defense is the control of abnormal operation and detection of failures. The essential means are control, limiting and protection systems and other surveillance features. The successful performance of Level 2 provisions will bring the plant back to normal operating conditions as soon as possible. Features of Level 2 should come into play whenever a significant deviation from normal operation conditions occurs, implying insufficient safety provisions at Level 1 and the occurrence of a PIE. Monitoring and surveillance measures are typically associated with this level of defense. Level 2 incorporates inherent plant features, such as core stability and thermal inertia, and systems to detect and/or control anticipated operational occurrences, with account taken of phenomena capable of causing further deterioration in the plant status. The systems to mitigate the consequences of such operating occurrences are designed to meet reliability objectives according to specific criteria (such as redundancy, layout and qualification). Diagnostic tools and equipment, such as automatic control systems, may be provided to actuate corrective actions before reactor protection limits are reached. In the NTPBMR designs, the reactivity control and shutdown system (RCSS) consists of independent and diverse systems used to control the reactor during normal operation conditions and, when required, to place the reactor in the hot shutdown condition. The control system serves to keep the reactor within normal operating limits, with an independent safety system providing capability to shut down the reactor if normal operating limits are exceeded. The combined use of an additional diverse system provides for maintaining the reactor subcritical indefinitely in a cold condition. There are limits placed on the depth of insertion of control assemblies to ensure that a sufficient immediate shutdown margin is always available. These systems are supported by a strong negative temperature reactivity coefficient that acts as an effective provision to limit maximum temperature of the fuel.
The vessel system design addresses the requirement for limiting helium leakage at normal operation to (typically) not more than 10% of the helium inventory in the primary circuit per year. In order to monitor and control the state of the vessel system and implement the “leak-before-break” concept, instrumentation may be provided that permits the identification and characterization of defects to be made on-line. The establishment of limiting conditions for operation (LCOs) for process variables will ensure the fulfillment of design basis accident assumptions, keeping their consequences within prescribed limits. Ongoing surveillance of quality and compliance with the design assumptions by means of in-service inspection and periodic testing of systems and plant components is also necessary to detect any degradation of equipment and systems before it can affect the safety of the plant.
The objective-provisions tree for the safety function control of reactivity at Level 2, with the correspondent acceptance criteria being: to limit insertion of reactivity to minimize automatic trips, to keep variables within their operating ranges, and to shutdown the reactor, if necessary. The objective-provisions tree for the safety function core heat removal, with the acceptance criterion al Level 2 being:
• To restore the balance between the heat generated and the heat removed, in order to comply with the allowed temperature ranges on fuel and structures established for anticipated operational occurrences.
The safety function confinement of radioactive materials has its objective-provisions tree depicted at the Level 2 of defense. The corresponding acceptance criterion is:
• To keep the concentration of radionuclides in the reactor coolant system and inside containment below the limits established for anticipated operational occurrences.
5.4. APPLICATION OF LEVEL 3 OF DEFENSE IN DEPTH FOR THE NTPBMR
The objective for Level 3 of defense is the control of accidents within the design basis. The essential means are inherent and engineered safety features and accident procedures. In spite of provisions for prevention and control of abnormal occurrences (failure of Levels 1 and 2), accident conditions may occur. Inherent safety features and protection systems and, if needed, engineered safety features, are provided to prevent evolution toward severe plant conditions and to confine radioactive materials. The measures taken at this level are aimed at preventing fuel damage in particular. All the safety related features are designed on the basis of postulated accidents representing the limiting loads of sets of similar events. Typical postulated accidents are those originating in the plant, such as the breach of a pipe containing primary coolant (a loss of coolant accident) or loss of control of reactivity (e.g. control rod withdrawal). Design and operating procedures are aimed at maintaining the effectiveness of the barriers, especially the fuel coating, in the event of such postulated accidents. Inherent features as well as active or passive systems are used. In the short term all these LOD are actuated inherently or by the reactor protection system when needed. If engineered systems (active or passive) are implemented, to ensure them a high reliability, the following design principles are adhered to:
• Redundancy (single failure criterion);
• Prevention of common mode failure due to internal or external hazards, by physical or spatial separation and structural protection;
• Prevention of common mode failure due to design, manufacturing, construction, commissioning, maintenance or other human intervention, by diversity or functional redundancy;
• Automation to reduce vulnerability to human failure, at least in the initial phase of an incident or an accident;
• Testability to provide clear evidence of LOD availability and performance;
• Qualification of LOD for specific environmental conditions that may result from an accident or an external hazard.
The fundamental safety concept for the NTPBMR designs is aimed at achieving a plant that has no physical process that could cause a radiation induced hazard outside the site boundary. This is mainly achieved by demonstrating that the heat loss from the reactor vessel ultimately exceeds the decay heat production in the post-accident condition and that the peak temperature reached in the core during the transient is below the demonstrated fuel degradation point and below the temperature at which the structures are affected. This is intended to preclude any prospect of significant core damage accident. Heat removal from the vessel is to be achieved by passive means.
The main provisions for Level 3 of defense in depth are:
• Decay heat removal during accidents by means of passive heat transport mechanisms (heat conduction, radiation, natural convection) to simple surface coolers. Besides dissipating the heat from the reactor cavity during normal operation, including shutdown, the Reactor Cavity Cooling System (RCCS) removes the decay heat during a loss of normal heat transfer functions (loss of coolant, loss of forced cooling). The objective is to prevent the reactor vessel, attachments, supports, instrumentation and the concrete walls from exceeding their design temperature limits. Natural processes, including thermal radiation, conduction and convection, are relied upon to transport the heat from the reactor vessel walls (with adequate emissivity) to the cooling panels of the RCCS;
• The strong negative reactivity temperature coefficient, in concert with large fuel power and temperature margins, provides a reliable inherent defense against positive reactivity insertion. The Reactivity Control and Shutdown System (RCSS), consisting of two independent and diverse systems, seismically designed and operated under a single failure criterion mode, provides further defense against reactivity events;
• The passive safety characteristics of the core (negative temperature and high temperature resistance) does not require an intact primary coolant pressure boundary (pipe break) to prevent significant core degradation. The fact that the fuel elements will be extracted from the reactor vessel in case of a loss of coolant event will have a significant effect on the objective-provisions trees, respectively to the safety functions control of reactivity, core heat removal and confinement of radioactive materials. The correspondent acceptance criteria are compatible with the objective for Level 3 of defense, which is the control of accidents within the design basis.
• The objective-provisions tree for the safety function control of reactivity at Level 3, with the correspondent acceptance criteria being: to limit the consequences of the maximum postulated insertion rate and amount of reactivity into the core, and to achieve and maintain adequate shutdown conditions.
• The objective-provisions tree for the safety function core heat removal, with the acceptance criterion at Level 3 being: adequate cooling of the fuel, vessel internals, vessel and reactor cavity by active/passive systems, via heat transfer to ultimate heat sink(s), ensuring core geometry, and reactor pressure vessel integrity.
• The safety function confinement of radioactive materials has its objective-provisions tree depicted at the Level 3 of defense. The corresponding acceptance criterion is: concentration of radionuclides (including fission products) below the limits established for design basis accident in the reactor coolant system and inside the reactor building, releases to the environment below the limits established for design basis accidents.
5.5. APPLICATION OF LEVEL 4 OF DEFENSE IN DEPTH FOR THE NTPBMR
The objective for Level 4 of defense is the control of severe plant conditions, including prevention of accident progression and mitigation of the consequences of severe accidents. The essential means are complementary measures and accident management. As noted earlier, severe plant conditions for the NTPBMR do not necessarily involve large releases from the fuel, as is generally understood to be the case for existing reactors in “severe accident” conditions. Since fuel melting is eliminated for all practical purposes, severe plant conditions are taken to be conditions that if left unmanaged could challenge the structural integrity of the core and thus the ability to analyze the course of an event. For the concept of defense in depth, it is assumed that the measures considered at the first three levels will ensure maintenance of structural integrity of the core and limit the potential radiation hazards for members of the public. Nevertheless, additional efforts, if deemed necessary, are made to further reduce the risk and consequences.
Accident management is not intended to be used to excuse design deficiencies at prior levels. The aim of the fourth level of defense is so to ensure that the plant safety related architecture is able to keep the consequences of the considered severe plant conditions within the allowable radiological limits. Consideration is given to severe plant conditions that were not explicitly addressed in the design (insufficient provisions at Levels 1 to 3) owing to their very low probabilities. Such plant conditions may be caused by multiple failures or by an extremely unlikely event such as a severe earthquake. Some of these conditions (e.g. large air ingress condition) bear a potential that radioactive materials could be released to the environment. The large thermal inertia of the plant and characteristics of the fuel and reactor internal structures will provide considerable time to deal with these conditions. If necessary, additional measures and procedures may be provided. Ancillary and support systems, if employed, would be designed, manufactured, constructed and maintained consistent with the required reliability. Measures for accident management are also aimed at controlling the course of severe plant conditions and mitigating their consequences.
Essential objectives of accident management are:
• To monitor the plant status;
• To maintain core sub criticality
• To protect the integrity of the coated fuel particles by ensuring heat removal from the core and preventing excessive loading conditions (both thermo-mechanical and chemical);
• To limit the release of radioactive material to the environment;
• To regain and maintain control of the plant.
The most important objective for mitigation of the consequences of an accident in Level 4 is the protection, to the maximum extent, of the capability of the coated fuel particles. The safety assessment shall also demonstrate that there is no risk for cliff edge effects. For this purpose some sequences must be excluded by design or practically excluded. The methodology to achieve such a demonstration must be defined to retain fission products. Inherent features are utilized where possible to attain this objective. Specific measures for accident management would be established on the basis of safety analysis and research results. These measures could utilize existing plant capabilities, including available non-safety classified equipment if they are operable for the accident conditions. Adequate staff preparation and training for such conditions is a prerequisite for effective accident management.
5.6. CONSIDERATIONS FOR LEVEL 5 OF DEFENSE IN DEPTH FOR NTPBMR
The objective for Level 5 of defense depth is mitigation of radiological consequences of significant releases of radioactive materials. The essential means are the off-site emergency response. Even if the efforts described in the foregoing are expected to be effective in limiting the consequences of severe plant conditions, it would be inconsistent with defense in depth to dismiss off-site emergency plans completely. These plans cover the functions of collecting and assessing information about the levels of exposures expected to occur in such very unlikely conditions, and the protective actions that could constitute intervention. The responsible authorities take the corresponding actions on the advice of the operating organization and the regulatory body. The extent of the emergency response plan should be commensurate with the radiological consequences predicted for the accident sequences which have been identified in the safety analysis. The aim is to design a plant for which sheltering and evacuation measures are not necessary.
SECTION SIX: SUMMARY AND COMPARISON OF NTPBMR AND LWR
The top-down approach discussed in this document is intended to be a general method for assessing the safety and developing safety requirements for the design of nuclear reactors taking into consideration the implementation of the principles of defense in depth. The method is applicable to any kind of reactor, however, how defense in depth is implemented and the implications on safety requirements are concept specific.
The application to the NTPBMR, although very preliminary, proved that the method is viable and useful. The specific features of the NTPBMR concept are significantly different from those of LWRs and they have great influence on the implementation of defense in depth. Stronger provisions at Level 1 could reduce the requirements on monitoring and controlling at Level 2. Passive safety features at Level 3 reduce the requirements on active engineered safety features at the same level and enhance the overall safety performance. Design basis accidents are mostly dealt with by inherent features and passive systems. The large thermal inertia of the NTPBMR enhances the effectiveness of Levels 2 and 3 of defense by providing very long times for systems and operator response and implementation of any mitigating measures. Needs for functional redundancies must be checked carefully through a deep and comprehensive analysis of the safety related architecture performance and reliability (PSA, for instance).
Quantitative comparisons between the safety performance of the NTPBMR and LWRs could show that similar postulated initiating events could lead to accident sequences with lower consequences or probabilities of occurrence in the NTPBMR than in LWRs. Of comparable importance is the potential that monitoring and surveillance requirements in Level 2 could be simplified or reduced in scope while providing an equivalent level of safety. The exercise also showed that areas such as the definition of success criteria for each Level of defense in depth and the integration of deterministic and probabilistic approaches need more investigation.
The NTPBMR fuel characteristics indicate that for internal initiating events, severe accident scenarios involving core melt can be practically excluded although severe scenarios involving extensive oxidation of the fuel could be envisaged. Design provisions and accident management measures must be considered carefully for very unlikely external challenges, like severe earthquakes, floods or airplane crashes in the context of retaining the structural integrity of the core.
6.2. COMPARISON BETWEEN THE SAFETY FEATURES OF WATER REACTORS AND THOSE OF NTPBMR
As stated earlier, in the recommended process the requirements for a specific type of reactor (e.g. NTPBMR) are to be generated through a critical interpretation of the ‘objectives’ and ‘essential means’ associated with each level of defense in depth for the reactors upon which the existing requirements are based. This requires a full understanding of the safety characteristics and features of the specific type of reactor under consideration as well as those of the water reactors on which the existing requirements were based. The applicability of an existing requirement must then be determined by a comparative evaluation of the two types of reactors. The purpose of this section of the document is to provide a summary comparison of the NTPBMR safety characteristics and features with those of water reactors.
Nothing presented here should be interpreted as criticism of the safety case for existing water reactors. Their performance speaks for itself as they have demonstrated a very high level of safety over many thousands of reactors years of operation. The water reactor safety case has been based upon highly reliable active systems to maintain the system parameters within the required envelope at all times and to respond rapidly as warranted for specific event conditions. The NTPBMR safety case utilizes the characteristics of ceramic fuel and core materials in conjunction with a passive design approach to avoid reliance on active systems, and thus takes a very different approach. Some of the key differences are summarized below, with representative characteristics from existing reactors upon which the current requirements are based referred to as the base case.
6.2.1. ALLOWED CONDITIONS WITHIN THE DESIGN BASIS
In the base case, fuel failure in the form of limited melting and/or cladding failure is allowed within the design basis, as long as a core geometry that is capable of being cooled is maintained. A comparison of allowed conditions within the design basis between the base case and the NTPBMR:
For the LWR case the combined effect of the initiating event and system response is allowed to fail two radionuclide containment boundaries in several events for the base case. For the NTPBMR design however the practice has been to preclude failure of any barrier except that associated with the initiating event. This difference, in conjunction with the ineffectiveness of a typical LWR containment design for the NTPBMR conditions requires a different approach to radionuclide containment than has been used in the base case.
6.2.2. FUEL FAILURE MECHANISMS
A comparison of fuel failure mechanisms for the base case and the NTPBMR shows that while there are a comparable number of mechanisms for fuel failure, there is a major difference with regard to the implications for safety, particularly with regard to the need for protection systems. Many of the mechanisms in the base case can cause fuel failure in the short term (seconds to minutes after the allowed envelope is exceeded). This leads to safety requirements for maintaining the allowed operating envelope on a moment-to-moment basis, and requirements for immediate response of mitigation systems. The last two mechanisms (clad ballooning/bursting and zirconium/water reaction) relate to loss of coolant accident conditions, where the residual fission and short term decay heat distribution is important and thus power level and power distribution just prior to the event are the primary operational state variables of interest. In the case of the NTPBMR, the failure mechanisms are for the most part related to long term operational conditions of the fuel. In the case of a loss of coolant, the slow response results in reaching a peak temperature days after the initiation of the event, thus operational state variables in the short term prior to the event are of little importance from a safety standpoint. A.3. CORE
6.2.3. TEMPERATURE MARGINS AND THERMAL RESPONSE
The combined effects of large temperature margins and slow thermal response are a central safety aspect of the NTPBMR, allowing major simplification of the operational safety requirements. In the base case, the structural limit is taken as the onset of rapid oxidation of the zircaloy cladding at approximately 1200ºC. The onset of cladding ballooning and bursting in the base case typically occurs at a lower temperature, but is a function of design specific internal pressurization and cladding mechanical properties. The chemical decomposition of silicon carbide, which becomes important around 2200ºC was taken as the structural limit for the NTPBMR fuel. As with the base case, other mechanisms such as diffusion of some fission products through the coatings begins at lower temperatures, affected by coating properties and fuel operating history. Note that the average fuel temperature in the reference case is higher than for the NTPBMR even though the coolant temperature is considerably lower. This is due to the much higher power densities in the base case, which cause a large temperature rise across the cladding/fuel pellet gap and within the fuel pellet.
The fuel melting temperature includes an allowance for a reduction of the melting point with fuel burnup. In the base case the design peak fuel temperature at full power is seen to be relatively close to the centerline melting limit. As with the fuel average temperature, the fuel maximum temperature is much higher for the base case due to the much higher power density. Protecting against this limit for the base case requires monitoring of power level and power distribution on a momentary basis. The large margins for the NTPBMR are a primary reason that fuel melting is not a credible condition.
The full power core adiabatic heatup rate is a hypothetical figure of merit for comparing thermal response. It is the rate of increase in temperature that would occur if the reactor core were operated at full power with no heat removal. The heatup rates for the base case are higher by between one and two orders of magnitude, depending on whether the coolant is present or not. The absence of the coolant has no significant effect on the response of the NTPBMR. This difference, which is the combined result of the low power density and high heat capacity of the NTPBMR, translates into a very slow response to conditions involving a mismatch between heat generation and removal.
6.2.4. OPERATIONAL IMPLICATIONS OF SLOW THERMAL RESPONSE
The long slow thermal response and large thermal margins of the NTPBMR opens the possibility of major simplifications of operational safety requirements. For example, the peak temperature on a loss of coolant with sustained loss of active cooling systems is reached days after the initiation of the event. The temperature distributions of interest with regard to fuel performance during the event are determined by the heat transfer characteristics for heat removal through the walls of the reactor vessel, and thus are effectively decoupled from the temperature distribution in the core prior to the event.
In addition, the power distribution affecting the event temperature distribution is the distribution of the longterm decay heat. This is determined by the long term core power distribution, and not significantly affected by the short term core power distribution prior to the event. The short term decay heat levels are important to the safety case for existing water reactors. In this case levels approach equilibrium within minutes to hours of operation prior to shutdown, thus the decay power distribution is strongly influenced by the short term power distribution prior to shutdown. For the long term decay heat of importance to the NTPBMR safety case, days to weeks of operation are required to approach equilibrium, and the shortterm power distribution prior to shutdown has no significant effect. The factors discussed above, in conjunction with the importance of controlling the integrated temperature, fluence and burnup history of the fuel, point to potential for a major simplification of operational safety requirements relative to existing water reactor plants. It is important to maintain the longterm core power and temperature distributions within an allowed envelope, but shortterm operational variations may be shown to be of no safety significance. Thus time compensated control and protection systems with restrictive requirements on accuracy and response times and resulting surveillance requirements typical of existing water reactors may be eliminated.