Alternative configurations for the QUADRISO fuel design concept

This letter extends the work performed at Argonne National Laboratory on the QUADRISO fuel particles concept for High Temperature Reactors. Different configurations of the QUADRISO fuel particles concept are proposed and examined. The concept of QUADRISO fuel particles allocates an extra burnable poison layer next to the fuel kernel to reduce the initial excess of reactivity and enhance the reactor performance. The alternative proposed configurations introduced in this letter compare the performances of the previous configuration with two alternative configurations where the burnable poison is mixed in the fuel kernel for simplifying the manufacture process or in the outer pyrocarbon layer.     

Effects of irradiation and design basis accident conditions on thermal properties of epoxy coating system for nuclear power plant

In this work, the thermal properties of epoxy coating system on carbon steel liner in an atomic reactor container have been investigated in terms of irradiation and design basis accident (DBA) conditions. Two epoxy coating systems were irradiated by different dose rates, i.e. 1×10⁶ and 5×10⁵ rad/h in a total dosage of 2×10⁸ rad/h. And, DBA tests were applied to them based on the specification. The glass transition temperature (T_g) and thermal stability of the epoxy coating systems after tests were measured by differential scanning calorimetry (DSC) and thermogravimetric analysis (TGA), respectively. As a result, the irradiation led to the fracture of internal structure in cured epoxy systems, resulting in significantly decreasing the thermal stability, as well as, the T_g. Also, the higher dose rate made a relaxation of epoxy resins to reduce the hardening of the network system. Whereas, the DBA test made an important role in the thermal stability in the epoxy coating systems, resulting in post-curing of the uncured epoxy resins. Consequently, it could be found that the thermal properties of the epoxy coating system were greatly influenced on the irradiation and DBA tests, which were probably due to the modification of internal network structure of the epoxy resins.     

Intensely irradiated steel components: Plastic and fracture properties, and a new concept of structural design criteria for assuring the structural integrity

In order to develop a systematic and reasonable concept assuring the structural integrity of components under intense neutron irradiation, two basic tensile properties, true stress-true strain (TS-TS) curves and fracture strain, were investigated on an austenitic stainless steel and martensitic steel. Application of Swift equation is confirmed to a large plastic strain range of TS-TS curves. Fracture strain ?_f data were well correlated as ?_f + ?₀ = const. where ?₀ is the pre-strain representing the irradiation hardening.Based on those formulations and available experimental information, several critical issues to be dealt with in developing the concept were identified possible reduction in ductility, significant change in mechanical properties, remarkable cyclic softening and other unique cyclic properties observed during a high-cycle fatigue testing, and the redundancy of the plastic collapse concept to bending. Existing structural codes are all based on the assumption that there will be no significant changes in mechanical properties during operation, and of high ductility. Therefore, a new concept for assuring structural integrity is required for application not only to components with high ductility but also components with reduced ductility. First, potential failure modes were identified, and a new and systematic concept was proposed for preventing these modes of failure, introducing a new concept of categorizing the loadings by stability of deformation process to fracture (as type F and M loadings). Based on the basic concept, a detailed concept of how to protect against ductile fracture was given, and loading type-dependent limiting parameters were set.Finally, application of the detailed concept was presented, especially on determination of loading type (in numerical approach, the formulation of TS-TS curves and fracture strain derived above are needed), and on how to determine the limiting parameters as allowable limits. Experiments were done to identify the loading type of a tensile loading acting on a structure with a discontinuity. Tensile loadings acting on an intensely neutron-irradiated flat plate with a hole in the center cause plastic tensile instability and necking at the minimum ligament section but do not initiate any surface crack at the initiation of necking.     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Dried hot concrete vessel for nuclear reactors: Proposal of a new design concept

The paper discusses special features of the concept of a dried hot prestressed concrete reactor vessel for pool-type liquid-metal cooled fast breeder reactors. In particular, the potentially advantageous feature of removing evaporable water from concrete when it is being kept hot is discussed along with its technological implications.     

Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

Spanish Breeding Blanket Technology Programme TECNO_FUS is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in [1]. The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau [2] and implemented in the OpenFOAM CFD toolbox [3]. The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 °C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed.Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid–solid coupled domain analysis.Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 °C, the required FCI material should have a very small effective heat transfer coefficient ((k/δ) ≤ 1 W/m²K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.     

The properties of WWER-1000 type materials obtained on the basis of a surveillance program

The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 units designated unit-1, -2 and -3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. Dependence of the radiation behavior of WWER-1000 type RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed.     

Determination of the plastic anisotropy or irradiated zircalloy tubes

Moscow Institute of Physical Investigations (MIFI). Translated from Atomnaya Énergiya, Vol. 71, No. 2, pp. 134–138, August, 1991.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

Physics design of initial and approach to equilibrium cores of a reactor concept for thorium utilization

A thermal reactor concept ‘a thorium breeder reactor’ (ATBR) was conceived and reported by the authors during 1998. The distinctive physical characteristics of ATBR core with different types of seed fuels have been discussed in subsequent publications. The equilibrium core of ATBR with Pu seed was shown to exhibit a flat and low excess reactivity for a fuel cycle duration of two years. Notably this is achieved by no conventional burnable poison but by intrinsic balancing of reactivity between fissile and fertile zones. In this paper we present the design of the initial core and the refueling strategy for subsequent fuel cycles to enable a smooth transition to the equilibrium core. Three fuel types with characteristics similar to the three batch fuels of equilibrium core were designed for the initial core. Fuel requirement for the initial core is 4673 kg of reactor grade (RG) Pu for a cycle length of two years at 1875 MWt as against the 2200 kg needed for each fuel cycle of equilibrium core for same quantum of energy. The core reactivity variation during the first fuel cycle is monotonic fall and is relatively high (∼40 mk) but gradually diminishes to ±5 mk for fuel cycles 5–8.     

Determination of floor response spectra on the basis of the response spectrum method

In this paper a method to determine floor response sepctra is proposed which is based on the modal analysis of a support structure with interaction-free, one-degree-of-freedom system attached. The time-consuming methods using real or artificial soil accelerations are avoided as well as some of the arbitrarinesses in the approaches of Biggs or Kapur-Shao.     

Second generation waste package design and storage concept for the Yucca Mountain Repository

The reference waste package design and operating mode to be used in the Yucca Mountain Repository is reviewed. An alternate (second generation) operating concept and waste package design is proposed to reduce the risk of localized corrosion of waste packages and to reduce repository costs. The second generation waste package design and storage concept is proposed for implementation after the initial licensing and operation of the reference repository design. Implementation of the second generation concept at Yucca Mountain would follow regulatory processes analogous to those used successfully to extend the design life and uprate the power of commercial light water nuclear reactors in the United States. The second generation concept utilizes the benefits of hot dry storage to minimize the potential for localized corrosion of the waste package by liquid electrolytes. The second generation concept permits major reductions in repository costs by increasing the number of fuel assemblies stored in each waste package, by eliminating the need for titanium drip shields and by fabricating the outer container from corrosion resistant low alloy carbon steel.     

Influence of mis-match of weld and base material creep properties on elevated temperature design of pressure vessels and piping

The stress and strain concentrations developed at the weldments during the long time operation of pressure vessels and piping at high temperature due to the mis-match in the creep properties of weldment constituents (weld, heat affected zone and base metal) are estimated using detailed finite element analysis. Three materials, viz. 2.25Cr 1Mo, SS 316 LN and modified 9Cr 1 Mo which are the most commonly used materials in the nuclear and thermal power plants are considered. A longitudinal seam weld with single and double V (X) configurations are analysed. Parametric studies have been done on weld angle and stresses. Based on the analysis, critical locations and the maximum stress concentration factors in the weldments for the above materials are identified. The weld design procedures of the currently used pressure vessel and piping codes are commented. The importance of ductility based failure criteria is emphasised.     

放射性物品运输容器防脆性断裂安全设计方法探讨

本文主要介绍了基于断裂力学的运输容器防脆性断裂安全设计,重点研究了防脆断条件中应力强度因子的计算方法。经分析,可以得出如下结论:RCC-M第Ⅰ卷附录ZG中的应力强度因子计算考虑了应力非线性分布和塑性区的影响,考虑因素比较全面,推荐采用。     

Determination of spatial resolution of plastic scintillation fiber array with a simple method

The spatial resolution of a position sensitive gamma-ray detector configuration based on plastic scintillation fiber array was measured using a Monte Carlo simulation method. Both point spread function and modulation transfer function （MTF） were presented. The factors that influence the spatial resolution were also discussed. The results of the simulation showed that the intrinsic spatial resolution was consistent with the size of the physical pixels and a few centimeters spatial resolution could be obtained under certain circumstances.     

New data on the geology and economics of the raw material basis of uranium in capitalist countries

In recent years, considerable data indicating important changes in the structure of the reserves and production of uranium in capitalist countries, the relative industrial importance of the various genetic types of uranium deposits, and the role of regional geological structures and stratigraphical systems have appeared in foreign literature, particularly in reports presented to the Second International Conference on the Peaceful Use of Atomic Energy (Geneva, 1958). These data lead to a new approach to the prospecting of uranium deposits and the development of new prospecting criteria.