Reinforced concrete containment wall elements — development of strain and failure in tested specimens

Research conducted by the NRC Office of Nuclear Regulatory Research in structural mechanics for reactor technology is reported in this paper. This research concerns applications to mechanical and structural systems and includes materials engineering related to light-water reactor nuclear power plants. The period between the Fifth International Conference on Structural Mechanics for Reactor Technology, held in Berlin in 1979, and the 1981 Sixth Conference in Paris is discussed. A listing of research projects recently completed and the status of research presently under way are included.     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

Containment response and radiological release for a TMLB′ accident sequence in a large dry containment

An analysis has been performed for the Bellefonte Pressurized Water Reactor (PWR) Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis, which include the effects of direct heating on containment loading and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating, which involves more than about 50% of the core, may fail the Bellefonte containment, but natural convection in the Reactor Coolant System (RCS) may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach, due to natural circulation, and after vessel breach, due to reevolution of retained fission products by fission product heating of RCS structures.     

Experimental and numerical approaches of aerosol removal in spray conditions for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear pressurized water reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety code validation. The present work is devoted to study a water spray injection used as a mitigation means in order to washout aerosol fission products.     

URANUS — A computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given.     

A simplified technique for shakedown limit load determination of a large square plate with a small central hole under cyclic biaxial loading

A simplified technique for determining the shakedown limit load of a structure was previously developed and successfully applied to benchmark shakedown problems involving uniaxial states of stress ( [Abdalla et al., 2007a], [Abdalla et al., 2007b] and [Abdalla et al., 2007c]). In this paper, the simplified technique is further developed to handle cyclic biaxial loading resulting in multi-axial states of stress within the large square plate with a small central hole problem. Two material models are adopted namely: an elastic-linear strain hardening material model obeying Ziegler's linear kinematic hardening (KH) rule and an elastic-perfectly-plastic (EPP) material model. The simplified technique utilizes the finite element (FE) method and employs small displacement formulation to determine the shakedown limit load without performing lengthy time consuming full elastic-plastic cyclic loading FE simulations or conventional iterative elastic techniques. The simplified technique is utilized to generate the shakedown domain for the plate problem subjected to cyclic biaxial tension along its edges. The outcomes of the simplified technique showed very good correlation with the results of analytical solutions as well as full elastic-plastic cyclic loading FE simulations. Material hardening showed no effect on the shakedown domain of the plate in comparison to employing EPP-material.     

Review of liquid metal corrosion issues for potential containment materials for liquid lead and lead–bismuth eutectic spallation targets as a neutron source

Lead (Pb) and lead–bismuth eutectic (44Pb–56Bi) have been the two primary candidate liquid metal target materials for the production of spallation neutrons. Selection of a container material for the liquid metal target will greatly affect the lifetime and safety of the target subsystem. For the liquid lead target, niobium–1 wt% zirconium (Nb–1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. In this paper, the oxidation rate of Nb–1Zr was studied based on the calculations of thickness loss resulting from oxidation. According to these calculations, it appeared that uncoated Nb–1Zr may be used for a 1-year operation at 900°C at P_O2=1×10^–6 Torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb–1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the liquid lead–bismuth eutectic target, three candidate containment materials are suggested, based on a literature survey of the materials’ compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr–1Mo, and 12Cr–1Mo (HT-9) steel. These materials seem to be used only if the lead–bismuth is thoroughly deoxidized and treated with zirconium and magnesium.     

Design and setup of a testing device to investigate a reduced sized attachment system mock up for the ITER EU HCPB-TBM under different mechanical loading conditions

In order to determine the forces acting on the EU-Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during operation, a measurement system is developed. Therefore, two force reconstruction (FR) methods using measured strain signals are selected that are suitable for the application to the TBM. The first one, the augmented Kalman filter is a combined deterministic-stochastic approach. A second FR method based on the concept of a model predictive controller is proposed in this paper, which uses an optimization algorithm. In order to test the selected methods a testing device has been built which can be used to apply different force excitations on a reduced sized TBM mock up and measure the resulting strain signals of 16 strain gages. A simple tube mock up has been designed and manufactured to test and calibrate the FR algorithms. In addition, a second TBM mock up with attachment system is described. Finally, first results of the FR of a worst-case test case from simulated strain data of the simple tube mock up are presented.     

Transient dynamic and inelastic analysis of shells of revolution — a survey of programs

Advances in the limits of structural use in the areospace and nuclear power industries over the past years have increased the requirements upon the applicable analytical computer programs to include accurate capabilities for inelastic and transient dynamic analyses. In many minds, however, this advanced capability is unequivocally linked with the large scale, general purpose, finite element programs. This idea is also combined with the view that such analyses are therefore prohibitively expensive and should be relegated to the “last resort” classification. While this, in the general sense, may indeed be the case, if the user needs only to analyze structures falling into limited categories, however, he may find that a variety of smaller special purpose programs are available which do not put an undue strain upon his resources. One such structural category is shells of revolution.This survey of programs will concentrate upon the analytical tools which have been developed predominantly for shells of revolution. The survey will be subdivided into three parts: (a) consideration of programs for transient dynamic analysis; (b) consideration of programs for inelastic analysis and finally; (c) consideration of programs capable of dynamic plasticity analysis. In each part, programs based upon finite difference, finite element, and numerical integration methods will be considered. The programs will be compared on the basis of analytical capabilities, and ease of idealization and use. In each part of the survey sample problems will be utilized to exemplify the state-of-the-art.     

事故条件下模块化小堆与同厂址大堆的辐射影响对比分析

本文针对海南小堆的实际厂址环境特征,根据机组初步的二级PSA源项,利用更实际的CALPUFF烟团模式开展事故条件下小堆和大堆对场外公众的辐射影响分析,比较不同事故下对周边居民和工作人员的受照特征。按照针对小堆的剂量准则,确定各种天气条件下满足该准则的距离,有助于更深入地认识小堆的事故特征及应急计划区划分等问题,为相关工程实践和应急监管工作提供参考。     

A study on the local behaviour of steel-concrete interfaces at and around large openings in the PSC inner containment dome

A finite element (FEM) study has been made to know the overall structural behaviour of a PSC inner containment (IC) dome having large steam generator (SG) openings with emphasis on the local behaviour of the steel-concrete interfaces near SG openings, under initial prestress transfer. The primary thrust of the work has been in the objective of predicting the possibilities of separation at the steel-concrete interface zones adjacent to the embedded plates of the SG openings. For the FEM analysis, the interface zone has been modeled using gap elements, the properties of which are derived from the results of the past experiments conducted on steel plate-concrete interface specimens. Important observations have been made regarding dome deformation and the stresses with special emphasis on the local behaviour of steel-concrete interfaces at and around SG openings.     

Study on kinetics of fast subcritical assembly—Contribution of delayed neutrons to the transition after a reactivity insertion in a subcritical system—

This study is an investigation of the effect of the delay neutron on the kinetics in the subcritical system. And, it proposes a method necessary for the kinetics code development that uses the Monte Carlo (MC) computation.

It is generally difficult to analyze three dimensional space and time dependent kinetics by using a MC method. It is because the sampling of the neutron in a region becomes difficult when conditions of the region changes with time. In this study, we consider about the effect of delayed neutron in the kinetics of ADS. The behavior of neutrons is considered spontaneous in this system. It means a neutron is absorbed or leaks in a short period, while the conditions of region do not change. Therefore they are treated by steady state calculation. On the other hand the densities of delayed neutron precursors changes slowly, and the conditions of region change. In the concept of developed MC method, the neutrons are calculated by using steady state equation at each time point, and the delayed neutron precursors are calculated by using time dependent equation. We tried to inspect the accuracy of this method by using a point equation. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as the solution where both neutrons and delayed neutron precursors are treated by using static equations. The obtained results show a good agreement between Φ1 and Φ*, though the Φ2 agrees with Φ* poorly in all cases. Especially, we showed that this technique was effective from the reactivity change by ADS, and the relation of a delayed neutron. Finally, the effect of the delay neutron on the beam trip in the neutron source for the drive was examined by using the technique of Φ2.     

Finite element and reliability: a method for compound variables—application on a cracked heating system

Nuclear engineering systems are designed to ensure safety criteria. To predict the behavior of mechanical systems, the finite element analysis (FEA) is actually the main tool for numerical analysis of mechanical problems. In order to design a system under data variability considerations, performance functions have to be defined by the relationship between the action effects and the material strengths. Then a certain level of safety should be satisfied with sufficiently high probability. This is the subject of the reliability theory. Controlling a FEA software in order to carry out the reliability analysis, it is to define a ‘combination method’. This paper proposes a general method for the reliability analysis combined with FEA codes. The method is efficient for independent, correlated and compound random variables. The proposed method is illustrated by numerical example of a cracked membrane exposed to thermal shortening. The risk to evaluate is represented by the crack propagation in the material.     

Experimental investigation in the TOSQAN facility of heat and mass transfers in a spray for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear-pressurized water reactor (PWR) containment. The TOSQAN facility which is highly instrumented with non-intrusive optical diagnostics is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we perform detailed characterization of the two-phase flow.