Progress of a cost optimization for an HLW repository in Korea

This paper provides the results of a cost optimization for a CANDU spent fuel canister as well as the operational duration of an HLW repository. From the design change of an advanced-CANDU spent fuel canister, the overall costs were expected to be reduced by 124 MEUR in the case of disposing of 36,000 tU in an HLW repository, and it was also found that the optimal operational duration for an HLW repository was 83 years, to minimize the total cost. But this operational duration was only calculated from the aspect of cost benefits with economics' perspectives.We confirmed that the canister and operational duration are the dominant cost drivers for surface facilities and underground facilities for a cost optimization, respectively. Especially, the manufacturing method of an outer canister using the cold spray coating technique which was developed through collaboration with a domestic company is suggested to minimize the overall costs.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 °C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.     

Cost efficiency analysis of disposal alternatives in Korea

This research presents the results of calculating the disposal cost efficiency for the four disposal alternatives for the CANDU spent fuel that are under development in Korea currently. The KRS-1 alternative, developed first, was set as the standard, and the efficiency of the KRS-1 alternative was assumed to be 100%.The cost calculation result shows that the A-KRS-22, which was developed most recently among the CANDU spent fuel disposal alternatives, manifested −61.7%, −45.7%, −47.0%, −78.9% and −61.7% when compared to the KRS-1 alternative concerning disposal tunnel excavation, disposal hole excavation, bentonite, disposal canister and backfilling.Moreover, the cost calculation method for the dominant cost driver that uses the unit disposal module concept for the calculation of cost efficiency was used. As for the reason that the standard for efficiency measurement was taken per each bundle, it is because the amount of bundle capacity concerning the spent fuel differs by disposal canister.     

Finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel storage cask

A spent fuel storage cask is required to prove the safety of its canister under a hypothetical accidental drop condition which means that the canister is assumed to be free dropped on to a pad of the storage cask during the loading of the canister into a storage cask. In this paper, finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel dry storage cask were carried out to improve the structural integrity of the canister under a hypothetical accidental drop condition. The pad of the storage cask was originally designed as cylindrical steel structure filled with concrete. The pad was modified by using the structure composed of steel and polyurethane-foam instead of the quarter of the upper concrete as an impact limiter. The effects of the shape and the thickness of the steel structure and the density of the polyurethane-foam which was used in between steel structures were studied. As the optimized pad of a spent fuel dry storage cask, the quarter of the upper concrete was replaced with 12 mm thick circular steel structure and polyurethane-foam whose density was 85 kg/m³. The drop tests of a 1/3 scale model for the canister on to the original pad and the optimized pad were conducted. The effect of the pad structure was evaluated from the drop tests. The optimized pad has a greater shock-absorbing effect than the original pad. In order to verify the analysis results, strains and accelerations in the time domain by the analytical methods were compared with those by a test. The numerical method of simulating the free drop test for a dry storage cask was verified and the numerical results were found to be reliable.     

The effect of radiation on the anaerobic corrosion of steel

To ensure the safe encapsulation of spent nuclear fuel elements for geological disposal, SKB of Sweden are considering using a canister, which consists of an outer copper canister and a cast iron insert. Previous work has investigated the rate of gas generation due to the anaerobic corrosion of ferrous materials over a range of conditions. This paper examines the effect of radiation on the corrosion of steel in repository environments. Tests were carried out at two temperatures (30 °C and 50 °C), two dose rates (11 Gray h⁻¹ and 300 Gray h⁻¹) and in two different artificial groundwaters, for exposure periods of several months. Radiation was found to enhance the corrosion rate at both dose rates but the greatest enhancement occurred at the higher dose rate. The corrosion products were predominantly magnetite, with some indications of unidentified higher oxidation state corrosion products being formed at the higher dose rates.     

Development of the detecting method of helium gas leak from canister

In the concrete cask storage system, spent fuel is installed and weld-sealed in a cylindrical container called a canister. The canister is filled with helium gas and its containment shall be maintained and inspected during storage. The helium gas enhances heat removal from spent fuel. When the helium gas leaks, the effect of helium gas convection is weakened in the canister. Thereof, the temperature on the canister surface changes.In present tests, it was found that temperatures of the center of the top and the bottom on the canister surface change remarkably during the helium gas leak. Therefore, we defined the temperature difference as ΔT_BT. And one can detect helium gas leak using the change of ΔT_BT. ΔT_BT increases monotonously toward a constant value during helium gas leak, even if the inlet air temperature drops. The helium gas leak can be detected at the early stage of the leak by observing both ΔT_BT and inlet air temperature.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

Criticality calculations of spent fuel in deep geological repository

Criticality calculations have been performed for a typical spent fuel disposal canister model filled with PWR fuel elements. Geometric and material properties of the disposal canister and disposal holes were modeled based on the Swedish preliminary disposal concept. Direct disposal of 5% enriched 16 × 16 PWR fuel was considered. We performed the calculations of the neutron multiplication factor using various disposal configurations, depending on the initial enrichment, fuel burnup, canister geometry and disposal holes configuration. The results showed that under normal conditions, when the canister is filled with fresh spent nuclear fuel, the system is deeply sub-critical. If it is assumed that the canister is faulty, leaking and filled with ground water, the system may become critical in the case of fresh fuel.     

一种干式外圆机械切割装置的研制

在乏燃料后处理中，需要回取已封装在乏燃料贮存容器中的乏燃料。根据热室使用环境及乏燃料贮存容器的特点，从耐辐射设计、乏燃料贮存容器固定、切割进给、切割刀具及刀具更换、放射性废物最少化等方面进行设计响应，研制了一种在热室内开启乏燃料贮存容器的干式外圆机械切割装置。功能性试验验证了该装置满足设计和使用要求。      

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

乏燃料密封容器开盖及其内容物回取技术研究

对于采用干湿法贮存的乏燃料而言,其后处理时面临的最大问题是如何安全高效地将乏燃料等内容物从封焊的密封容器中取出。针对这一问题,基于乏燃料密封容器及其内容物的结构特点,开展了乏燃料密封容器开盖及内容物回取技术研究,综合考虑切割热室使用环境、内容物回取后的收集和转移以及产生废物的收集处理等因素,制定了合理可行的开盖及回取工艺,研制了用于开盖和筒体分段切割的解体装置以及回取和吊装工具,并通过试验验证了工艺的可行性以及研制的工装具的可用性。      

Validation of the thermal calculation method and the applicability of confinement monitoring with a full-scale mock-up test of a concrete cask

Abstract

The Mitsui Engineering & Shipbuilding Co. Ltd (MES) has designed and fabricated a full-scale mock-up system that can be used to store spent nuclear fuel (SNF). The system is made up of two parts; a concrete shield that is vented and an inner steel canister that provides containment of the SNF. A benchmark analysis of this storage system was carried out using a combined thermal calculation method. Initially airflows and temperatures outside the canister were calculated using a three-dimensional thermal flow analysis method. The results from this analysis were used as the boundary conditions to calculate the maximum temperatures inside the canister using a two-dimensional heat transfer method. The calculated results agreed well with the measurements and the validity of the combined method of analysis was confirmed. Since all measured temperatures were within their acceptable limits, it was also confirmed that the concrete cask storage system has sufficient heat removal capability. MES has also proposed a new canister confinement monitoring system. It is based on the relationship between the inner pressure of the canister and the temperature of the canister lid and the pedestal. The validity and applicability of the system are confirmed by the full-scale mock-up experiment results. The conceptual design of the monitoring system is considered, and the system can realised at low cost, with high reliability and easy maintenance.     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

The Swedish nuclear waste program and the long-term corrosion behaviour of copper

The principal strategy for high-level radioactive waste disposal in Sweden is to enclose the spent fuel in tightly sealed copper canisters that are embedded in bentonite clay about 500 m down in the Swedish bedrock. Besides rock movements, the biggest threat to the canister in the repository is corrosion. ‘Nature’ has proven that copper can last many million of years under proper conditions, bentonite clay has existed for many million years, and the Fennoscandia bedrock shield is stable. The groundwater may not stay the very same over very long periods considering glaciations, but this will not have dramatic consequences for the canister performance. While nature has shown the way, research refines and verifies. The most important task from a corrosion perspective is to ascertain a proper near-field environment. The background and status of the Swedish nuclear waste program are presented together with information about the long-term corrosion behaviour of copper with focus on the oxic period.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Investigation of 85Kr Emission and Its Effect on the Radiation Conditions during the Preparation of BOR-60 Fuel for Recovery

The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base at the Scientific-Research Institute of Nuclear Reactors, are examined. The ⁸⁵Kr emission at the stages of preparation of the spent BOR-60 fuel assemblies for recovery is determined experimentally. It is shown that the maximum ⁸⁵Kr emission as a result of destruction of fuel element cladding with oxide uranium fuel is 68%; its contribution to the irradiation dose to the public as a result of mechanical disassembly of the fuel elements in a single BOR-60 fuel assembly with 10% burnup and a 10-yr holding time does not exceed 1·10^–4% of the dose limit (1 mSv/yr).     

Sensitivity analysis of fission product activity in primary coolant of typical PWRs

A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in primary coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power (P), failed fuel fraction (D), coolant leakage rate (L), total mass of coolant (m) and the let-down flow rate (Q) have been calculated and the values: 1.0, 0.857, −2.0177 × 10⁻⁶, 2.349 × 10⁻⁴, −2.329 × 10⁻⁴ have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.     

Performance of Fuel Failure Detection System for Coated Particle Fuels

An experimental system was developed for a study of fuel failure detection (FFD) method for coated particle fuels (CPF's) of a high-temperature gas-cooled reactor. Various performances of the FFD-system were examined using a CPF-irradiation rig in the Japan Material Testing Reactor. By experiments, it was made sure that the counting rates of fission products (FP's), released from the CPF's, change with the reactor-power and the fuel-temperature remarkably even during the normal reactor operation. Also, an ability of the selective detection of only short-life FP-nuclides was studied in relation to the travelling time of the sampling gas. The results showed that the contributions of the short-life FP-nuclides such as Kr-89 and Kr-90 are more than 80 percent to the total FP-counting rate at the shortest travelling time of 120 sec. It is concluded that the selective detection of only the short-life FP-nuclides can be realized by controlling the travelling time properly.     

Thermo-mechanical FE-analysis of residual stresses and stress redistribution in butt welding of a copper canister for spent nuclear fuel

The transient and residual temperature, stress and strain field present during electron beam welding of a plane copper end to a canister for spent nuclear fuel is calculated by the use of FEM. The subsequent stress redistribution is calculated up to 10,000 years. The canister consists of two concentric cylinders, an inner steel cylinder containing the spent nuclear fuel and an outer copper cylinder. It was found that the maximum plastic strain (plastic+creep) accumulated in the (possibly brittle) heat affected zone is ≈7%, which seems to be well below the reported ductility for the copper used.