Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li₂TiO₃) and beryllium–titanium (Be₁₂Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.     

A fluid-structure finite element method for the analysis of reactor safety problems

A method is presented for the safety analysis of reactor containment structures by means of finite elements. The finite element equations of both fluid and structural elements for arbitrarily large, non-linear response are developed and the way in which they are combined is indicated. Both explicit and implicit integration of the equations in time is considered. Three examples of the application of these methods to the analyses of reactor safety problems are described.     

Classification method of severe accident condition for the development of severe accident instrumentation and monitoring system in nuclear power plant

Instrumentation and monitoring systems in a nuclear power plant are very important to monitor plant conditions for safe operations and a plant shutdown. The severe accident at TOKYO ELECTRIC POWER COMPANY's Fukushima Daiichi Nuclear Power Station (hereinafter called as TF1) in March 2011 caused several severe situations such as core damage, hydrogen explosion, etc. Lessons learned from the severe accident at TF1 show that an appropriate operable instrumentation and monitoring system for a severe accident should be developed so that the system will deliver an appropriate performance for mitigation of severe accident condition in a nuclear power plant.

This paper proposes the classification method of severe accident condition for the development of an appropriate operable instrumentation and monitoring system for a severe accident based on the problem analysis of monitoring variables during the severe accident at TF1. The classification is formed on the basis of the integrity of boundary for plant safety and the successful (or unsuccessful) condition of the cooling water injection, and is used for an establishment of defining severe accident environmental conditions for the instrumentation and monitoring system. Examples of the establishment method are also shown in this paper.     

Human reliability analysis for operators in the digital main control rooms of nuclear power plants

ABSTRACT

As the main control room of nuclear power plants (NPPs) has been gradually digitized, new human reliability problems may emerge because of a series of new changes in the cognitive processes, behavioral patterns, and error mechanisms of operators. Aiming to address this situation, this paper proposes a method as guidance for human reliability analysis (HRA) of different cognitive Stages. This method first constructs the influencing factors of three cognitive processes, including monitoring, decision-making, and execution of actions, and then evaluates the weights of these influencing factors through an analytic hierarchy process (AHP). In this study, the parameters used in the proposed HRA method were determined by analyzing the test data obtained from a simulation model, and the results demonstrated the rationality and feasibility of the proposed method. A case example using this HRA method was given in which the human error probabilities at three stages in a nuclear power plant (NPP) steam generator tube ruptures (SGTR) accident were obtained. In summary, the proposed method is a simple and feasible HRA tool that can be applied in digital NPP main control rooms (MCRs).     

A fluid–solid finite element method for the analysis of reactor safety problems

The work presented herein can broadly be categorized as a fluid–structure interaction problem. The response of a circular cylindrical structure subjected to cross flow is examined using the finite element method for both the liquid and the structure domains. The cylindrical tube is mounted elastically at the ends and is free to move under the action of the unsteady flow-induced forces. The fluid is considered to be acoustic compressible and viscous. A Galerkin finite element method implemented on a triangular mesh is used to solve the time-dependent Navier–Stokes equations. The cylinder motion is modeled using a five-degrees of freedom generalized shell element structural dynamics model. The numerical simulations of the response of the calandria tubes/pressure tubes, adjustor rod and shut-off rod of a nuclear reactor are presented. A few typical results are presented to assess the accuracy and applicability of the developed modules.     

Evaluation method for structural integrity assessment in core disruptive accident of fast reactor

A procedure to evaluate the structural integrity of a reactor vessel under CDA condition has been developed based on simulative experiments under CDA loading using 1/20- and 1/10-scale steel models of the demonstration FBR in Japan (DFBR). For the experiments, a constant pressure source has been developed using a gas accumulator. Simulation analyses were also conducted using the dynamic analysis program AUTODYN-2D, which can simulate fluid-structure interaction and non-linear material behaviour in two-dimensional geometry. The actual dynamic material properties, whose strain rates were from 0.1 to 10 s⁻¹, were used in the analyses. The pressure history and the dynamic response of structures were compared between experimental results and the analysis, and the appropriateness of the numerical analytical method for the CDA problem was confirmed. In both experimental and numerical approaches, the effect of internal structures has been investigated, and a drastic strain reduction was observed owing to the energy absorption by internal structures. Furthermore, material tests and element tests including weld joint tests and burst tests were carried out to define a rationalised strain limit under multi-axial dynamic loading condition. Using the developed evaluation procedure, a preliminary evaluation for DFBR was carried out and the robustness of the reactor boundary in DFBR against CDA loading was also confirmed.     

Quantifying reactor safety margins Part 6: A physically based method of estimating pwr large break loss of coolant accident pct

The Nuclear Regulatory Commission (NRC) has issued a revised Emergency Core Cooling System (ECCS) rule allowing the use of best estimate computer codes for safety analysis. To support this revised rule the NRC and its consultants and contractors have developed and demonstrated the Code Scaling. Applicability and Uncertainty (CSAU) Evaluation Methodology. This effort lead to increased understanding of the phenomena, and their relative dominance, during Large Break Loss of Coolant Accidents (LBLOCAs) in Pressurized Water Reactors (PWRs). Consequently, it became possible, as is done in this paper, to develop a method for establishing clad temperature history by using physically based arguments and engineering correlations. The results from this method are compared with similar uncertainty estimates based on large computer codes. These comparisons provide a rationale, based on physical arguments, for evaluating the large computer code based estimates of uncertainty.     

A new physics design of control safety rods for prototype fast breeder reactor

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B₄C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B₄C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B₄C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B₄C sections in top and bottom. There is significant savings in the initial inventory of enriched B₄C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles.     

Design of engineered barriers for filled emplacement rooms in the planned Konrad repository

In the Federal Republic of Germany it is intended to dispose of radioactive waste with negligible heat impact on the host rock in the former Konrad iron ore mine. The waste packages will be disposed of in emplacement rooms which, once completely filled, will be sealed by closing-off structures. The requirements to be imposed on the closing-off structures are derived by a safety analysis for the operational phase. Referring to different levels of requirements two concepts of closing-off structures are introduced. The feasibility of the almost tight closing-off structure is shown in situ.     

A probability approach to estimating the safety of a reactor installation

Krasnaya Zvezda Scientific-Designer Bureau. Central Institute of Atomic Energy, All-Union Research Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 74, No. 1, pp. 38–42, January, 1993.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part II. Testing the fuel

This paper summarizes what is done for the experimental testing of cermet fuel with various matrix materials. Low neutron absorption, high heat conductivity, good corrosion resistance in water, low chemical interaction with cladding (zirconium alloy) and UO₂ in normal and accident conditions, technological ability — are the requirements of the matrix material [1]. Suitability of the proposed solutions to the cermet fuel design with respect to these requirements was proven through a series of experiments simulating fuel operating and accidental conditions.     

Two viewpoints for software failures and their relation in probabilistic safety assessment of digital instrumentation and control systems

As the use of digital systems in nuclear power plants increases, the reliability of the software becomes one of the important issues in probabilistic safety assessment. In this paper, two viewpoints for a software failure during the operation of a digital system or a statistical software test are identified, and the relation between them is provided. In conventional software reliability analysis, a failure is mainly viewed with respect to the system operation. A new viewpoint with respect to the system input is suggested. The failure probability density functions for the two viewpoints are defined, and the relation between the two failure probability density functions is derived. Each failure probability density function can be derived from the other failure probability density function by applying the derived relation between the two failure probability density functions. The usefulness of the derived relation is demonstrated by applying it to the failure data obtained from the software testing of a real system. The two viewpoints and their relation, as identified in this paper, are expected to help us extend our understanding of the reliability of safety-critical software.     

A comparative method for improving the reliability of brittle components

Calculating the absolute reliability built in a product is often an extremely difficult task because of the complexity of the physical processes and physical mechanisms underlying the failure modes, the complex influence of the environment and the operational loads, the variability associated with reliability-critical design parameters and the non-robustness of the prediction models. Predicting the probability of failure of loaded components with complex shape for example is associated with uncertainty related to: the type of existing flaws initiating fracture, the size distributions of the flaws, the locations and the orientations of the flaws and the microstructure and its local properties. Capturing these types of uncertainty, necessary for a correct prediction of the reliability of components is a formidable task which does not need to be addressed if a comparative reliability method is employed, especially if the focus is on reliability improvement. The new comparative method for improving the resistance to failure initiated by flaws proposed here is based on an assumed failure criterion, an equation linking the probability that a flaw will be critical with the probability of failure associated with the component and a finite element solution for the distribution of the principal stresses in the loaded component. The probability that a flaw will be critical is determined directly, after a finite number of steps equal to the number of finite elements into which the component is divided. An advantage of the proposed comparative method for improving the resistance to failure initiated by flaws is that it does not rely on a Monte Carlo simulation and does not depend on knowledge of the size distribution of the flaws and the material properties. This essentially eliminates uncertainty associated with the material properties and the population of flaws.On the basis of a theoretical analysis we also show that, contrary to the common belief, in general, for non-interacting flaws randomly located in a stressed volume, the distribution of the minimum failure stress is not necessarily described by a Weibull distribution. For the simple case of a single group of flaws all of which become critical beyond a particular threshold value for example, the Weibull distribution fails to predict correctly the probability of failure. If in a particular load range, no new critical flaws are created by increasing the applied stress, the Weibull distribution also fails to predict correctly the probability of failure of the component. In these cases however, the probability of failure is correctly predicted by the suggested alternative equation. The suggested equation is the correct mathematical formulation of the weakest-link concept related to random flaws in a stressed volume. The equation does not require any assumption concerning the physical nature of the flaws and the physical mechanism of failure and can be applied in any situation of locally initiated failure by non-interacting entities.     

Improved Monte Carlo-perturbation method for estimation of control rod worths in a research reactor

A hybrid method dedicated to improve the experimental technique for estimation of control rod worths in a research reactor is presented. The method uses a combination of Monte Carlo technique and perturbation theory. Perturbation method is used to obtain the equation for the relative efficiency of control rod insertion. A series of coefficients, describing the axial absorption profile are used to correct the equation for a composite rod, having a complicated burn-up irradiation history. These coefficients have to be determined – by experiment or by using some theoretical/numerical method. In the present paper they are derived from the macroscopic absorption cross-sections, obtained from detailed Monte Carlo calculations by MCNPX 2.6.F of the axial burn-up profile during control rod life. The method is validated on measurements of control rod worths at the BR2 reactor. Comparison with direct MCNPX evaluations of control rod worths is also presented.     

A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

A severe accident in a marine nuclear reactor leads to radionuclide leakage, which causes hidden dangers to workers and has adverse effects of environmental pollution. It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor. Therefore, this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe...     

A multibarrier safety concept for the reactor pressure vessel of the Stade Nuclear Power Plant

Two specific problems within the safety case of Stade RPV have been analysed: brittle fracture initiation and arrest under strip type emergency core cooling conditions and safety margins against ductile failure from deep cracks as postulated by ASME- and German KTA-rules. For EOL material conditions exclusion of initiation is shown for cracks of more than twice the size which is safely detectable by NDE; for arbitrarily postulated large cracks it is demonstrated that they are arrested well within the allowed depth of of the wall thickness; therefore no critical crack size exists for Stade RPV under strip cooling. Growth in depth of an assumed circumferential flaw in the girth weld embrittled at EOL could occur only at upper shelf temperatures and by loads higher than about twice the service pressure; leak before break was demonstrated in a constraint-modified J − R-curve crack-growth analysis. But neither a transient nor the plant itself would be able to provide the necessary high loads. The LEFM and EPFM proofs are summarized in a multibarrier safety scheme.