Drop testing conducted to benchmark the Shippingport reactor pressure vessel package safety analysis

The decommissioned Shippingport reactor pressure vessel and its integral neutron shield tank were transported from Shippingport, Pennsylvania, via barge to Richland, Washington, for burial in the Hanford Site radioactive waste disposal area. To ensure that the reactor pressure vessel/neutron shield tank assembly could be shipped safely without undue risk to the public or the environment, the reactor pressure vessel/neutron shield tank assembly was certified by the U.S. Department of Energy as a type B package. A safety analysis report for packaging was prepared in accordance with U.S. Department of Energy requirements to provide the technical basis for the U.S. Department of Energy certification. The reactor pressure vessel/neutron shield tank package is a monolithic structure of lightweight concrete and steel. Its estimated weight is 844 t (930 tons). To substantiate multidimensional inelastic analyses, a series of 11 drop tests was conducted on 7 benchmark models from heights of 30.5 cm (1 ft), 9.14 m (30 ft), and 13.7 m (45 ft). Technical evaluation and correlation of the test data were performed in conjunction with the structural analysis and assessment of the package. This paper provides a comprehensive discussion on the benchmark drop models and specific drop tests and also addresses the results obtained from comparing technical data with analytical data.     

Monte Carlo modeling of KAMINI

Monte Carlo modeling of the Kalpakkam Mini Reactor (KAMINI) has been carried out for the first time by using Monte Carlo code (MCNP4A) and continuous energy cross-sections. The safety control plate (SCP) drop experiment is simulated and the computed integral worth of the SCPs is compared with the measured value. The measured axial neutron flux profile and foil reaction rates in one of the in-core irradiation location and the foil reaction rates at the west beam port are also compared with the predicted results. The agreement between measurements and calculations is quite satisfactory. It is confirmed from the calculation and measurement that the north thimble is having nearly 10–20% higher neutron flux as compared to the south thimble depending on the exact elevation.     

Refined Monte Carlo analysis of the H.B. Robinson-2 reactor pressure vessel dosimetry benchmark

Refined analysis, based on use of the Monte Carlo code MCNPX-2.4.0, is presented for the “H.B. Robinson-2 pressure vessel dosimetry benchmark”, which is a part of the Radiation Shielding and Dosimetry Experiments Database (SINBAD). First, the performance of the Monte Carlo methodology is reassessed relative to the reported deterministic results obtained with DORT. Thereby, the analysis is accompanied by a quantitative evaluation of the optimal energy cut-off value for each of the in- and ex-vessel dosimeters that were employed. Second, a more realistic definition of the neutron source is implemented than proposed in the benchmark. Thus, the current procedure for power-to-neutron-source-strength conversion, as also for explicitly considering the burnup-dependent fuel assembly-wise average fission neutron spectrum, is found to affect the calculated values significantly.     

基于Hoogenboom基准模型的SuperMC全堆芯计算能力校验

基于国际经典的压水堆全堆芯Hoogenboom基准模型,对超级蒙特卡罗核计算仿真软件系统SuperMC(Super Monte Carlo Simulation Program for Nuclear and Radiation Process)进行了校验。对有效增殖因数keff、功率等反应堆关键参数进行了计算与正确性校验,对并行效率进行了分析。结果显示,SuperMC的计算结果与MCNP(Monte Carlo N Particle Transport Code)吻合较好,在使用640核计算时并行效率高达98.7%,初步验证了SuperMC在全堆芯计算中的准确性及高效性。     

Analysis of the IAEA research reactor benchmark problem by the RETRAC-PC code

A basic approach to perform safety analysis of a nuclear research reactor consists in using deterministic methods to verify that the established acceptance criteria related to fuel integrity are fulfilled during all the stages of the facility lifetime. These methods should be validated against a large set of experimental and postulated transients. Since measured data are not easily available in the literature, the IAEA defined typical transients in a generic 10-MW MTR nuclear reactor core as a benchmark test for computational tools verification. In this framework, an assessment study of the coupled kinetic–thermal–hydraulic RETRAC-PC code is presented herein. The considered cases include the analysis of core dynamic under ramp positive reactivity insertion, and loss of flow transients. In general, the obtained results are satisfactory and agree with results obtained by other similar codes.     

Cross-section adjustment in the analysis of benchmark experiments

An iteration procedure is given for the solution of the exact non-linear least-squares equation for cross-section adjustment. The solution is made feasible by properly weighting all detectors used in the Benchmark experiment and combining them to one ‘global’ detector. The adjoint fluxes and sensitivity coefficients must be calculated only for this one global detector so that in each iteration cycle the code ANISN and SWANLAKE have to be run only once.Simultaneously an expression was found for the error-covariance matrix of the adjusted cross-section set.     

Detail analysis of fusion neutronics benchmark experiment for iron

In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of ⁵⁷Fe in JENDL-3.3 caused the overestimation.     

A benchmark on the calculation of kinetic parameters based on reactivity effect experiments in the CROCUS reactor

Measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions. The experimental configurations were also calculated by four different calculation methods. For each of the supercritical configurations, the absolute reactivity value has been determined in two different ways, viz.: (i) through direct comparison of the multiplication factor obtained employing a given calculation method with the corresponding value for the critical case (calculated reactivity: ρ_calc); (ii) by application of the inhour equation using the kinetic parameters obtained for the critical configuration and the measured inverse reactor period (measured reactivity: ρ_meas). The calculated multiplication factors for the reference critical configuration, as well as ρ_calc for the supercritical cases, are found to be in good agreement. However, the values of ρ_meas produced by two of the applied calculation methods differ appreciably from the corresponding ρ_calc values, clearly indicating deficiencies in the kinetic parameters obtained from these methods.     

Validation of IRBURN calculation code system through burnup benchmark analysis

Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes.     

基于组件计算的燃耗实验基准题建模分析

组件计算在堆芯核设计中占有重要地位。组件程序的燃耗计算精度对核反应堆堆芯的功率分布、换料寿期及反应性控制设计方面具有重要意义。为了评估用于堆芯燃耗计算的多群常数库的准确性,本文运用DRAGON计算程序建立了燃耗实验计算模型,采用SFCOMPO-2.0燃耗实验基准题提供的乏燃料样品燃耗历史参数及最终核素组分信息,对Takahama-3反应堆、H.B. Robinson-2反应堆及Beznau-1反应堆系列样品进行了建模计算,并将计算结果与SFCOMPO-2.0数据库中的基准实验结果进行了对比和分析。结果表明:多数核素的模拟结果与基准值符合良好,误差在10%以内。同时本文对理论计算值与基准实验值存在差异较大的几种核素进行了相应讨论,并对样品计算结果进行了对比分析。     

Structural analysis of reactor fuel elements

An overview of fuel element modeling is presented that traces the development of codes for the prediction of light-water reactor and fast breeder reactor fuel element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations.Current efforts include modeling of alternate fuel systems, analysis of local fuel cladding interactions and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.     

Earthquake response analysis of reactor structures

The general nature of the principles upon which earthquake resistant design is based is described with particular reference to components and elements of nuclear reactor facilities. Special attention is paid to the response and design criteria of items of equipment or of components that are mounted on or attached to responding elements, and basic procedures are developed to bound the dynamic response of such items.

Consideration is given to vertical as well as horizontal excitation, and the combination of the effects of the various exciations. Suitable approximations are developed for inelastic response estimates.

One section of the paper is devoted to relative motions of points some distance apart, and to bounds for such relative motions.

Recommendations are made for the general criteria governing the design of nuclear facilities, including the basic parameters governing response characteristics and energy absorption.     

Benchmark analysis of SPERT IV reactor

Benchmark calculations have been performed for SPERT IV D-12/25 core. Experimental data of the core was provided by International Atomic Energy Agency (IAEA). Combination of WIMS/D4 and CITATION codes has been used for performing the neutronic analysis of the reactor. Lattice calculations have been performed through WIMS/D4 while 3-dimensional reactor core has been modeled in CITATION. Ten energy groups were considered for these calculations. Energy wise microscopic cross-sections were generated for fuel, control absorber, control follower, guide tube, grid plate, reflector and structural regions separately of the core using WIMS/D4. Thermal neutron flux profiles at different axial and radial locations of the core were evaluated. Critical position of the control rods, excess reactivity, shut down margin, control rod worth, reactivity feed back coefficients and kinetic parameters of the core were estimated. Reasonable agreement has been found between experimentally determined and the calculated parameters.     

Transient thermal-hydraulic analysis for reactor cores

A new computational method is presented for a transient, thermal-hydraulic, multichannel analysis. The method is developed based on the concept of artificial compressibility to preserve the elliptic character of the reactor core flow in order to satisfy the realistic pressure boundary conditions, and to account for the discontinuities of the emprical correlations simulating the flow resistances. The computer code (RETSAC) developed by implementing the method presented in this paper can be categorized as a fourth generation multichannel computer code. This new computer code has been compared with the widely used marching techniques, such as COBRA IIIC (the third generation). The numerical results clearly indicate the situations in which the marching technique may or may not be appropriate. Furthermore, the RETSAC computer code can calculate various normal or off-normal reactor core flows which the third generation codes could not handle without a substantial increase of computer time.     

Seismic analysis of heavy-liquid-metal-cooled reactor vessels

The linear-elastic seismic analysis for the vessel of a heavy-liquid-metal reactor was undertaken based on the Design Response Spectrum (DRS) approach and for a 0.5 g earthquake. Four support types for the vessel were analyzed and it was found that the roll support exhibits the best overall performance. The variation of the normal-mode frequency and total peak stress intensity with the vessel diameter and thickness was also studied. It was found that the frequency of the first normal mode increases with increasing vessel diameter and thickness, while the total peak stress intensity decreases with increasing vessel thickness and is roughly independent of the vessel diameter. Two new dimensionless groups are introduced that enable correlation of the frequency and stress-intensity data with adequate accuracy. It is proposed that these correlations be used for quick estimate of the seismic response of a vessel of arbitrary size, material and contained (heavy) fluid, subject to an earthquake of arbitrary magnitude.     

Seismic analysis of cracked reactor vessels

This paper deals with the dynamic response of a thin finite, elastic circular cylindrical shell representing a reactor vessel to time-dependent loadings symmetrical with respect to the axis of the cylinder. The shell contains an axial through-crack of length 2c. The dynamic counterpart of Donnell's shell equations are used in this investigation. Extensive numerical results are presented for stress intensity factors in aluminum and steel vessels and results are discussed.     

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.