INTOR — a first-generation tokamak experimental reactor

An intensive, year-long, international evaluation of the next major tokamak beyond the generation of large experiments currently under construction was carried out during 1979. This evaluation consisted of the definition of objectives an assessment of the physics and technology base and R&D needs and the identification of a set of parameters that physically characterize the machine.     

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method.     

An experimental study of inter-subassembly heat transfer during natural circulation decay heat removal in fast breeder reactors

The evaluation of core thermohydraulics under natural circulation conditions is significant to utilize passive safety features of fast breeder reactors (FBRs). Under low flow conditions, it is predicted that buoyancy effects and heat transfer through wrapper tubes, i.e. inter-subassembly heat transfer, will significantly influence the flow and temperature distributions in subassemblies. Thus, steady-state sodium experiments were carried out using a three-subassembly model. The transverse temperature distributions in the subassemblies were measured under conditions wherein inter-subassembly heat transfer occurred. A wall subchannel factor was introduced to estimate the sodium temperature near the wrapper wall, which characterizes the inter-subassembly heat transfer. This factor enables a one-dimensional system code to predict the inter-subassembly heat transfer accurately. The characteristics of the factor were studied experimentally. It was shown that a buoyancy parameter, Gr*/Re, and a heat flux ratio of wrapper wall to pin surface were essential to predict the wall subchannel factor and also the peaking factor. Experimental analyses were also carried out using a three-dimensional analysis code that modeled the multi-bundle system. Good agreement between experiments and calculations was obtained for temperature distributions influenced by the inter-subassembly heat transfer.     

Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and simplified refueling by replacing an IFA in which all the core and blanket fuel elements are comprised. An on-site storage cask achieves on-site decay heat removal of an IFA. After 3 years of on-site storage, an IFA together with an on-site storage cask can be transported directly to the reprocessing plant without any intermediate steps. Significant improvement of inherent safety features and plant availability has been discussed. Decay heat removal capability, safety consideration on criticality of the IFA and shielding design of the on-site storage cask has been confirmed. The RAPID refueling concept possesses high resistance to state-supported removal of plutonium for nuclear weapons production.     

聚变堆氚增殖层中子学分析

D-T聚变堆包层的主要功能包括氚增殖、能量转换射层蔽等,包层中子学设计的主要原则是满足聚变堆的氚自持,一般要求包层氚增殖比TBR＞1.1.使用与时间有关的扩散理论和本征函数展开方法,研究不同几何线度、6Li丰度的LI2O、LiPb包层材料14MeV源下的系统通量、氚增殖比影响,及在不同6Li丰度下6Li、7Li造氚随时间变化的规律.计算中使用了30群截面数据,微观数据来自ENDF/B-VI及JEF-2.2.     

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans.The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes.A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor.The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured.Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.     

Irradiation analysis of U, Am samples irradiated in experimental fast reactor “Joyo” for protected plutonium production in fast breeder reactor blanket

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper summarizes analysis of the individual Am and U samples irradiation in Joyo to re-evaluate the results of Pu isotopes in the measure of proliferation resistance, and to combine the results for the prediction of DU-Am irradiation especially in the production of Pu isotopes. By the prediction of DU-Am oxide fuel in fast reactor environment with detail fuel irradiation analysis, it was confirmed that neutron moderation and fuel size affects the produced Pu isotope and its vector due to the very high sensitivity of ²³⁸U resonance capture reaction, the larger diameter fuel is more preferable in the case of moderated neutron spectrum environment for denaturing Pu in fast reactor blanket. Finally proliferation resistance of all the Pu produced in U, Am sample irradiation and DU-Am fuel irradiation prediction were evaluated based on decay heat and spontaneous fission neutron rate, and it was confirmed ²⁴¹Am produces un-attractive Pu to abuse from the beginning to the end of irradiation, and more than 2% of ²⁴¹Am doping is required to enhance the proliferation resistance of Pu to MOX grade and Kessler’s proposal in moderated neutron spectrum environment in fast reactor.     

The He/He heat exchanger — Design and semitechnical testing

This is a report on the development of the He/He heat exchanger which is used for high-temperature reactors (HTR) combined with the steam gasification of coal. A concept has been agreed on the basis of the requirements resulting from the application of the HTR. Subsequently those steps, which are required for the development of this component up to construction maturity are described. Simultaneously, questions dealing with material, construction, design, manufacture and related experimental development are taken into consideration.     

The nuclear safety project at KfK 1972–1986 — Objectives and results

In 1972 the light water reactor safety activities conducted at the Karlsruhe Nuclear Research Center (KfK) were combined under the Nuclear Safety Project (PNS). Its primary objective was to assess in quantifiable terms the safety reserves which are provided in nuclear power plant design in a conservative approach. While in the initial phase R&D work conducted under the project was largely characterized by investigations of the design basis accidents, mainly the loss-of-coolant accident, emphasis in the past decade has been shifted more and more towards severe core and core meltdown accident analysis. The activities comprise both theoretical studies and experimental investigations, often performed in adequate, large-scale facilities. All activities have been an essential part of the reactor safety research program of the Federal Ministry for Research and Technology (BMFT) and have been coordinated with a number of other programs conducted in Germany and abroad. This paper gives a broad overview of PNS contributions to LWR safety research in the past 15 years and summarizes the results, comparing them with the general goals defined. In conclusion, the attempt is made to give an outlook on remaining activities in LWR safety research being carried out by KfK.     

The SCARABEE facility — Its main characteristics and the experimental program

The first phase of the SCARABEE programme has already been done with fresh fuel on single and seven pin bundles. We are describing the SCARABEE facility which includes the reactor, the sodium loop, the test section, the equipment for post mortem examination, the instrumentation and the recording systems.The methods used to determine the experimental parameters such as the power generated in the pins or the heat losses are presented.The different types and number of experiments are also described.     

Chromium-molybdenum steels for fusion reactor first walls — a review

Chromium-molybdenum steels have recently become of interest as a first wall and blanket structural material for fusion reactors. This application will be assessed, and possible approaches on how Cr---Mo alloys may be further developed for this application will be proposed.Generally, the Cr---Mo steels can be divided into two categories: unmodified, basically Cr---Mo---C steels and Cr---Mo---C steels modified by the addition of carbide-forming elements. Extensive research and development efforts have been conducted on the unmodified steels, especially Cr-1 Mo and 12 Cr---Mo steels. Considerable work has also been done on 12 Cr---Mo steels modified with additions of V, Nb, Ti and W. In recent years much of the research effort on this type of alloy has been directed at developing modified Cr---Mo steels with less than 12% Cr (generally 9%) for applications where the “stainless” properties imparted by chromium additions of at least 12% are not needed.We will examine the unmodified and modified steels in terms of hardenability, precipitation processes (stability at elevated temperatures), strength, and toughness. Where possible, we will discuss the effects of irradiation on these properties. Such a study leads to the types of tradeoffs that may be necessary between the well-researched unmodified Cr-1 Mo steel and a high-chromium modified steel.     

Thermal hydraulic characteristics analysis of direct residual heat removal system for China pressurized reactor

An innovative Direct Residual Heat Removal System (DRHRS) is proposed for Pressurized Water Reactor (PWR) in this paper. The new designed parallel DRHRS is different from traditional Passive Residual Heat Removal System (PRHRS), which is connected to steam generation. The thermal hydraulic transient analysis of the new designed DRHRS for CPR1000 has been carried out using the widely accepted safety analysis software RELAP5. The new designed DRHRS is directly connected to the primary loop, which consists of three independent parallel loops, three intermediate cooling circuits and an air loop. The transient behaviors of passive safety system are studied, and design parameter sensitivity analysis is carried out. Results show that during Station Black_Out (SBO) accident, natural circulations are established stably in passive safety system so that core decay is continuously removed from primary loop. And the new designed DRHRS has the capability of removing residual heat to the atmosphere without any external energy input at different surrounding environmental temperature. In emergency, the DRHRS directly remove core decay heat from reactor outlet, and efficiency of residual heat removal is improved. Moreover, reactor power plant maintains safe even if double-ended rupture of a single tube during SBO accident occurs. Thus, the designed DRHRS has great significance for increasing the degree of inherent safety features of CPR1000.     

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 °C which is substantially lower than ∼627 °C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a crucial factor for consideration in safety design. This study provides future researchers with a guideline on designing safety measures for the fourth generation of the fast reactors with no particular reference to any specific manufacturer.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

The inherent safety characteristics of an unconventional LMR design—The Trench Reactor

This paper presents the transient behavior during off-normal operation of an unconventional liquid metal reactor design, called the Trench Reactor. Under the postulated accident conditions, this reactor design responds in an inherently safe manner to loss of heat sink accidents, loss of flow accidents, overcooling accidents and transient overpower accidents with 25 cents of reactivity insertion. The characteristics that cause such inherently save behavior are the properties of the materials and the configuration of the reactor primary system, even without any activated safety devices.     

Static and stability analysis of the inner vessel of a fast breeder reactor

Liquid metal fast breeder nuclear reactors demand the usage of large sized thin shells for their reactor vessel components due to low operating pressure and high thermal load. Buckling is a very important aspect in the design of these vessels. In this article, analysis of the inner vessel of a typical 500 MWe fast breeder reactor is presented. Here, two different geometric configurations of the inner vessel are considered. One configuration is with the conical step joining the upper and the lower cylindrical portions, and the other is with the toroidal bottom joining the upper cylindrical part. The buckling strength of the vessel for both configurations are calculated and compared. Also, the effects of thermal load, initial geometric imperfection, geometric nonlinearity, etc. are investigated. The finite element method is used for analysis.     

Direct heat transfer using liquid metals — Problems and technical realization

This paper discusses the possibilities of using metal baths as heat-transfer systems for gas cooled high-temperature reactors under special consideration of the gasification of solid combustibles.     

NONSAP — A nonlinear structural analysis program

The current version of the computer program NONSAP for linear and nonlinear, static and dynamic finite element analysis is presented. The solution capabilities, the numerical techniques used, the finite element library, the logical construction of the program and storage allocations are discussed. The solutions of some sample problems considered during the development of the program are presented.