Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.     

Verification by calculations of shock-like loads on transport and storage casks

In this paper, criteria for the verification of numerical results for type B tests are proposed. Furthermore, some well-known commercially available FEM codes are investigated by analysing benchmark problems especially designed for transport packaging and by analysing drop tests experimentally performed on an original transport and storage cask. The results are checked by the proposed verification criteria and are found to be reliable.     

The interim fuel storage facility of the PBMR

The PBMR’s spent fuel and partially burnt fuel are stored in the sphere storage system (SSS), which acts as the interim fuel storage facility of the plant. It is unique in the world since the fuel is stored in bulk containers (called storage tanks), each capable of holding more than 500,000 spheres for a period of about 80 years. The SSS has the ability to transfer the contents of one tank to another tank, and to return partially burnt fuel back to the reactor core for re-fuelling.The storage tanks are individually sealed carbon steel pressure vessels. They form the final barrier of any fission products that have diffused from the fuel spheres. Sub-criticality is achieved through the geometric cross-section of the tank, and by taking credit for fuel burn-up. Protection from the corrosive environment where the PBMR Demonstration plant will be built is achieved by actively cooling the tank with clean dry air. In the event of an active cooling failure, louvres open automatically and cooling is done passively via natural convection making use of the chimney-effect. Sufficient radiation protection is provided around each tank to facilitate maintenance and inspection operations where needed.The design of the SSS is nearing the end of its basic design phase, and for some components, detail design work has already commenced. The design complies with all spent fuel storage requirements and is seen as a cost-effective solution for the interim storage of PBMR spent fuel.     

Investigation of seal effects according to axial compression variation of metal seals for transport and storage casks

Abstract

Casks for the transport and storage of heat generating radioactive waste in Germany are normally provided with screwed lid systems, which are in most cases equipped with double jacket metal seals with an inner spring wire to provide long term resistance to the seal compression force. Preservation of the high sealing quality of those seals under operational and accidental stress conditions is essentially important to the safety of those casks. Relative displacements of the lid system surfaces caused by specific impact scenarios cannot be excluded and have to be evaluated with respect to a possible increase in the leakage rate.

To get representative data for such metal sealed lid systems, BAM has developed a special conceptualised flange system placed in an appropriate testing machine for relevant mechanical loading of the metal seals under static and cyclic conditions. Furthermore, the flange system enables continuous measurement of the standard helium leakage rate during each test.

The primary aim of the investigation is to identify the correlation between variation of installation conditions (axial displacements) caused by external loads and the standard helium leakage rate. An essential parameter in this case is the useable resilience r_u of a metal seal under relevant stress conditions. The useable resilience r_u is the vertical difference in the cross-section between the seal's assembling status and the point where the leakage rate, by means of external load relieving, exceeds the quality criterion of 10^–8 Pa m³ s^–1. Load relieving can instantly occur due to modification of the seal groove dimension caused by accident impacts and deformation of the lid system. Furthermore, component specific basis data for the development of finite element calculation models should be collected. In the tests, seals are subjected to static and cyclic loads. All tests are performed at ambient temperature.

This paper presents the test configuration, different test series and results of the current experiments. Typical load–displacement–leakage rate correlations are presented and discussed.     

Analysis for seismic response of dry storage facility for spent fuel

Most of the dry storage systems for spent fuel are freestanding, which leads to stability concerns in an earthquake. In this study, as a safety check, the ABAQUS/Explicit code is adopted to analyse the seismic response of the dry storage facility planned to be installed at Nuclear Power Plant #1 (NPP1) in Taiwan. A 3D coupled finite element (FE) model was established, which consisted of a freestanding cask, a concrete pad, and underneath soils interacting with frictional contact interfaces. The scenario earthquake used in the model included an artificial earthquake compatible to the design spectrum of NPP1, and a strong ground motion modified from the time history recorded during the Chi-Chi earthquake. The results show that the freestanding cask will slide, but not tip over, during strong earthquakes. The scale of the sliding is very small and a collision between casks will not occur. In addition, the differential settlement of the foundation pad that takes place due to the weight of the casks increases the sliding potential of the casks during earthquakes.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

Cooldown testing of nuclear fuel casks

Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown.     

乏燃料干式贮存设施临界计算研究

以HI-STORM 100乏燃料干式贮存设施内部装载AFA-3G燃料组件为研究对象,用MCNP(Monte Carlo N Particle Transport Code)4C程序,通过改变贮存设施内外的水密度,采用新燃料假设对不同工况下的临界安全进行研究。结果表明,在正常工况下,keff远低于0.93,是临界安全的。在事故工况下,当水密度大于0.8 g·cm-3时,存在临界安全问题。然后选取适当的核素,通过使用ORIGEN-ARP程序,得到不同燃耗下核素的组成,在同一模型下考虑燃耗信任制,对干式贮存设施的临界安全进行研究。在此基础上,给出了乏燃料干式贮存设施临界安全工作的相关建议。     

Use of Alcan's Al-B4C metal matrix composites as neutron absorber material in TN International's transportation and storage casks

Abstract

Major issues in the area of transportation and/or storage of radioactive materials are reliability and safety of engineering components. Among the functions to be undertaken, transportation and storage systems shall allow the criticality control of the transported matter, the control of its temperature, as well as the capacity to withstand the mechanical stresses due to normal, incidental and accidental conditions of use. In most cases, criticality control requires the use of an internal arrangement made of a neutron absorber material, which must also have high thermal conductivity properties to ensure the temperature control. When, as in many AREVA-TN International applications, the design takes credit of the neutron absorber material as a structural component, it must show high mechanical performance. Alcan's Al-B₄C metal matrix composites (Al-B₄C MMCs) meet all the above mentioned requirements, due to their special capability of capturing neutrons, their light weight, and their superior thermal conductivity and mechanical properties. The significant advantage of Alcan's technology is its flexibility with regards to a wide range of boron carbide contents and matrix alloys (from AA1XXX to AA6XXX). This enables the adjustment of the properties to the exact needs of the design. TN International presently uses extruded and/or rolled Al-B₄C MMC parts in several of its internal arrangements. The present paper gives an overview of the manufacture processes of Alcan's Al-B₄C MMCs, from the mixing of B₄C into liquid aluminium to the extrusion and rolling operations. It describes the methods and results for the qualification tests in terms of the neutron absorption, thermal, physical and mechanical properties of the material. Finally, details are given on the use of Alcan's MMCs as a neutron absorber with enough credit for structural material in TN International's TN24 designs.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Chloride induced stress corrosion cracking of candidate canister materials for dry storage of spent fuel

Susceptibility to chloride induced stress corrosion cracking (ESCC) of candidate canister materials, UNS S31260 and UNS S31254 stainless steels (SS), was investigated by a constant load test in air at temperatures of 343 and 353 K with relative humidity (RH) of 35%, and at 373 K without controlling RH. UNS S31260 and UNS S31254 SS did not fail until 37,700 h at 353 K with RH = 35%, where UNS S30403 SS failed within 250–500 h. The same tendency also was obtained at 343 K, suggesting the superior ESCC resistance of UNS S31260 and UNS S31254 SS. Even rust was not observed on the specimens tested at the temperature of 373 K. To explain the higher ESCC resistance, the pitting potential was measured in the saturated synthetic sea water at temperatures from 303 to 353 K, since ESCC is usually associated with localized corrosion such as pitting and may be closely related to the corrosion resistance. The pitting potentials of UNS S31260 and UNS S31254 SS were much higher than that of UNS S30403 SS. Thus, it was concluded that the superior ESCC resistance is attributable to the higher resistance of UNS S31260 and UNS S31254 SS to pitting corrosion. The critical relative humidity for ESCC, under which no ESCC occurs, is equal to or higher than 15% at temperatures < 353 K judging from ESCC behavior of UNS S30400 SS.     

Internal cask content collisions during drop test of transport casks for radioactive materials

Abstract

In transport casks for radioactive materials, significantly large axial and radial gaps between cask and internal content are often present because of certain specific geometrical dimensions of the content (e.g. spent fuel elements) or thermal reasons. The possibility of inner relative movement between content and cask will increase if the content is not fixed. During drop testing, these movements can lead to internal cask content collisions, causing significantly high loads on the cask components and the content itself. Especially in vertical drop test orientations onto a lid side of the cask, an internal collision induced by a delayed impact of the content onto the inner side of the lid can cause high stress peaks in the lid and the lid bolts with the risk of component failure as well as impairment of the leak tightness of the closure system. This paper reflects causes and effects of the phenomenon of internal impact on the basis of experimental results obtained from instrumented drop tests with transport casks and on the basis of analytical approaches. Furthermore, the paper concludes the importance of consideration of possible cask content collisions in the safety analysis of transport casks for radioactive materials under accident conditions of transport.     

基于蒙特卡罗方法的新型乏燃料干式贮存容器辐射安全仿真验证

针对自主设计的贮存24组燃耗深度为45 GWD/MTU的乏燃料组件的CHN-24型专用容器临界及辐射屏蔽问题,采用蒙特卡罗程序MCNP,建立CHN-24容器临界及辐射屏蔽计算模型。研究结果表明:正常贮存条件下容器内乏燃料的有效增殖因数(k_(eff))为0.283,发生浸水事故时,k_(eff)随着容器内水位升高逐渐增大,注满水时keff达到最大值0.706;容器表面剂量当量率随浸水量增大而减小;正常贮存条件下,即无水浸入时,容器表面及距表面1 m处的最大剂量当量率值分别为0.42 m Sv·h~(-1)、0.08 m Sv·h~(-1)。以上均符合国际原子能机构规定的临界及剂量安全标准,同时表明蒙特卡罗方法可应用于乏燃料容器的临界及辐射屏蔽安全验证。该研究为我国研发具有自主知识产权的核电乏燃料贮存专用容器提供了一定的参考依据。     

Full-scope simulation of a dry storage cask using computational fluid dynamics

A computational fluid dynamics (CFD) analysis of a TN24P cask was performed through a full-scope simulation using FLUENT. In order to establish the analysis methodology while minimizing the computational burden, the sensitivities of various parameters were investigated by constructing a small-scale model. The full-scale CFD predictions of the TN24P cask were compared with the experimental data and COBRA-SFS results. There was good agreement between the FLUENT predictions and the experimental data. FLUENT showed similar temperature predictions to COBRA-SFS, while there were deviations between FLUENT and COBRA-SFS in the velocity predictions. By conducting sensitivity studies for the application uncertainties using a full-scale simulation, it was found that the basket gap size was the most sensitive parameter in the analysis.     

Thermal hydraulic analysis compared with tests of full-scale concrete casks

To develop a thermal analysis method for the concrete cask, numerical calculation based on thermal hydraulic phenomena was performed. In the present calculation model, calculation area was divided into two parts. One is inside of the canister and the other is outside of the canister. These two parts were combined at the surface of the canister. In the model of the outside, k– turbulence model was adopted for air flow region. Comparing calculation results with test results, it was found that the analysis method was valid for normal and accident conditions of the storage.     

Demonstration drop test and design enhancement of the CANDU spent fuel storage basket in dry storage facility

A dry interim storage facility has been constructed at the Wolsung power plant in Korea. This dry storage facility has seven separated modules. There are 40 long slender cylinders in one module. In one cylinder, ten baskets where sixty CANDU spent fuel bundles are loaded are stacked and stored. For this dry storage facility, analyses and tests for hypothetical accident conditions that might occur while moving and storing the baskets into a cylinder were performed. In a demonstration test, one of test basket models did not satisfy one of the safety-related requirements. Thus, the revised basket designs were generated using a structural evaluation based on finite element analyses and specimen tests. Among these revised designs, one design was chosen as a final revised basket design. The final revised design was the one creating the largest reduction of plastic strain in the upper welding region. It was also the one that required the smallest design change from the current basket design for easy adoption. Drop and leak tests were performed, and the final revised basket satisfied all performance requirements.     

上海光源储存环新注入方式

上海光源储存环的新工作模式和新装插入件会导致动力学孔径和注入效率的减小,传统的注入方式不再满足恒流注入要求。分别从改变注入方式和减小注入束流发射度两个角度解决注入对动力学孔径的需求。脉冲多极铁与凸轨相结合的注入方式可以将动力学孔径的要求降低到7 mm以下。利用斜四极铁在高能输运线上进行水平和垂直发射度互换可以将注入束流发射度降低到近1/10而降低注入点的束斑尺寸。两者都可以将注入效率提高到95%以上。二者在更小动力学孔径的情况下同时使用还可以减小脉冲多极铁的强度。     

Creep assessment of Zry-4 cladded high burnup fuel under dry storage

Cladding creep rupture is thought to be the most likely and limiting failure mechanism of spent fuel in dry storage. In spite of being highly unlikely, the current trend towards high burnups is drawing further attention to the potential creep effect on cladding integrity of fuels burnt over 45 GWd/tU.This paper explores the burnup influence on cladding creep during dry storage by modelling two different high burnup scenarios (51 GWd/tU and 67 GWd/tU). In addition, sensitivity of the results to the in-reactor average power and power history has been conducted. The computation tool used in this study has been an extension of FRAPCON-3.4 capable of simulating dry storage scenarios. Burnup and average linear power have been shown to make creep grow quite substantially during the first two years in dry storage, adopting a quasi-asymptotic trend from then on. However, even though this profile seems to have a generic nature, the net creep value reached depends not just on integral and average variables, but also on magnitudes describing the entire irradiation history, like linear power history. In none of the cases explored creep approaches the 1% threshold. In-reactor FGR modelling has been highlighted as a key element to get accurate estimates of creep.     

Basic principles for ensuring nuclear safety in designs of transportable packaged assemblies

Conclusions Any design of transportable packaged assemblies intended for the conveyance of spent nuclear fuel must meet the requirements of PBYa-06-88-77 and must give representative evidence of nuclear safety. The proposed classification of damage established a suitable procedure for calculating the critical parameters of the system when carrying out the preliminary analysis of the nuclear safety of packaged assemblies at different stages of their design. After the preliminary analysis of the nuclear safety of the TK-6 and TK-10 packaged units, the acceptable significant and dangerous damage has been established, which leads to displacement of the fuel within the package, which creates a real possibility for the technical assurance of nuclear safety.Translated from Atomnaya Énergiya, Vol. 49, No. 4, pp. 216–218, October, 1980.