Spent Pressurized Water Reactor Fuel Research Articles

Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

Fractional Release Behavior of Volatile and Semivolatile Fission Products During a Voloxidation and OREOX Treatment of Spent PWR Fuel

Quantitative analysis of the fission gas release characteristics during the voloxidation and oxidation and reduction of oxide fuel (OREOX) processes of spent pressurized water reactor (PWR) fuel was carried out by spent PWR fuel in a hot cell of the DUPIC Fuel Development Facility. The release characteristics of 85Kr and 14C fission gases during voloxidation process at 500°C are closely linked to the degree of conversion efficiency of UO2 to U3O8 powder, and it can be interpreted that the release from grain boundary would be dominated during this step. Volatile fission gases of 14C and 85Kr were released to near completion during the OREOX process. Both the 14C and 85Kr have similar release characteristics under the voloxidation and OREOX process conditions. A higher burnup spent fuel showed a higher release fraction than that of a low burnup fuel during the voloxidation step. It was also observed that the release fraction of semivolatile Cs was ~16% during a reduction at 1000°C of the oxidized powder, but over 90% during the voloxidation at 1250°C.

Automatic refueling simulation of a DUPIC fuel transition core

This paper presents an automatic refueling simulation of a 713MWe Canada deuterium uranium (CANDU-6) reactor transition to Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC) fuel system. The simulation searches for a set of refueling channels based on the zone controller unit water level, the reactivity insertion tilt in two axial zones and the discharge burnup. Compared to the natural uranium core, the transition core has two distinctive features such as an axial power peaking and an inconsistent reactivity insertion into two axial zones with a fueling direction, which are accommodated by adjusting the reference power of the central zone and complying with the reactivity insertion tilt in the axial direction. Starting from the natural uranium equilibrium core, it took 730 full power days to completely fill the core with the DUPIC fuel, and the characteristics of the transition core represented by the maximum channel and bundle powers were far below the license limits.

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Examination of Spent Pressurized Water Reactor Fuel Rods After 15 Years in Dry Storage

For [approximately equal to]15 yr Dominion Generation’s Surry Nuclear Station 15 × 15 Westinghouse pressurized water reactor (PWR) fuel was stored in a dry inert-atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory at peak cladding temperatures that decreased from ~350 to 150°C. Before storage, the loaded cask was subjected to thermal-benchmark tests, during which time the peak temperatures were greater than 400°C. The cask was opened to examine the fuel rods for degradation and to determine if they were suitable for extended storage. No fuel rod breaches and no visible degradation or crud/oxide spallation from the fuel rod surface were observed. The results from profilometry, gas release measurements, metallographic examinations, microhardness determination, and cladding hydrogen behavior are reported in this paper.It appears that little or no fission gas was released from the fuel pellets during either the thermal-benchmark tests or the long-term storage. In the central region of the fuel column, where the axial temperature gradient in storage is small, the measured hydrogen content in the cladding is consistent with the thickness of the oxide layer. At ~1 m above the fuel midplane, where a steep temperature gradient existed in the cask, less hydrogen is present than would be expected from the oxide thickness that developed in-reactor. Migration of hydrogen during dry storage probably occurred and may signal a higher-than-expected concentration at the cooler ends of the rod. The volume of hydrides varies azimuthally around the cladding, and at some elevations, the hydrides appear to have segregated somewhat to the inner and outer cladding surfaces. It is, however, impossible to determine if this segregation occurred in-reactor or during transportation, thermal-benchmark tests, or the dry storage period. The hydrides retained the circumferential orientation typical of prestorage PWR fuel rods. Little or no cladding creep occurred during thermal-benchmark testing and dry storage. It is anticipated that the creep would not increase significantly during additional storage because of the lower temperature after 15 yr, continual decrease in temperature from the reduction in decay heat, and concurrent reductions in internal rod pressure and stress. This paper describes the results of the characterization of the fuel and intact cladding, as well as the implications of these results for long-term (i.e., beyond 20 yr) dry-cask storage.

Determination of Ru, Rh, Pd, Te, Mo and Zr in spent pressurized water reactor fuels by ion exchange and extraction chromatographic separations and inductively coupled plasma atomic emission spectrometric analysis

A study has been carried out on the quantitative analysis of Ru, Rh, Pd, Te, Mo and Zr, which are main components of undissolved materials dispersed in nitric acid dissolver solutions of spent pressurized water reactor (PWR) fuels, using inductively coupled plasma atomic emission spectrometry (ICP-AES). The dissolution of the insoluble residues filtrated from the dissolver solutions by acid digestion bomb technique and the separation of Ru, Rh, Pd, Te, Mo and Zr from U and transuranous elements in the filtrate by ion exchange, and di(2-ethylhexyl) phosphoric acid (HDEHP) extraction chromatography were discussed in detail. The applicability and versatility of the proposed separation procedure combined with ICP-AES analysis were evaluated by chemically characterizing the spent PWR fuels with burnups from 36,000 to 39,000 MWd/MtU.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors—IV: DUPIC Fuel Cycle Cost

This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW·h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW·h. Considering the uncertainty (0.40 to 0.44 mills/kW·h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ~20% and reduce the spent fuel arising by ~65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors—I: DUPIC Fuel Fabrication Cost

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (~49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors—III: Spent DUPIC Fuel Disposal Cost

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Composition Heterogeneity Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)ȔI: Deterministic Analysis

The effect of DUPIC fuel composition heterogeneity on CANDU core performance was assessed for three candidate DUPIC fuel options: the fissile content adjustment method, reactivity control by slightly enriched and depleted uranium, and reactivity control by natural uranium. The fissile content adjustment method produces DUPIC fuel of fixed 235U and 239Pu contents, while the reactivity control method produces DUPIC fuel of uniform reactivity at the fresh condition. To assess the uncertainty of the core performance parameter associated with the isotopic variation, the sensitivity coefficients were generated by the generalized perturbation theory for the lattice parameter and zone controller level perturbations. The uncertainty was then estimated for three key core performance parameters: maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF). The fissile content adjustment method was shown to have a smaller uncertainty in the core performance parameter than with the reactivity control options. For the fissile content adjustment method, the average uncertainties of MCP, MBP, and CPPF were estimated to be 1.3, 2.5, and 1.2%, respectively, with 95% confidence level, when simulated for specified burnup points of the equilibrium core.

Composition Heterogeneity Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)—II: Statistical Analysis

The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: fissile content adjustment, reactivity control by slightly enriched and depleted uranium, and reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types, which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters, such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF), are close to those of the core that has a single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are <0.6, 1.5, and 0.8%, respectively. This study has shown that the three DUPIC fuel options reduce the composition heterogeneity effectively, and the zone power control system has a sufficient margin to adjust the perturbations caused by fuel composition heterogeneity.

Separation of fission products from spent pressurized water reactor fuels by anion exchange and extraction chromatography for inductively coupled plasma atomic emission spectrometric analysis

A study has been carried out on the separation of fission products from spent pressurized water reactor (PWR) fuels and their quantitative determination using inductively coupled plasma atomic emission spectrometry (ICP-AES). Plutonium and uranium were separated from fission products using anion exchange and tri- n-butylphosphate (TBP) extraction chromatography, respectively. Americium was separated and fission products such as Sr, Ba, Cd, La, Ce, Pr, Nd, Sm, Eu, Gd and Y were quantitatively recovered using di(2-ethylhexyl)phosphoric acid (HDEHP) extraction chromatography. An ICP-AES/shielding system, which was specially designed and built for the analysis of radioactive materials, was employed and the analytical precision for all metal elements studied was found to be <5%. Three spent PWR fuels whose burnup rates were between 15,000 and 35,000 MWd/MTU were analysed and then the relation between the burnup and the quantity of the fission products was compared to that calculated by burnup code, ORIGEN 2.

Reactivity Control Method for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for DUPIC fuel, which is made from spent pressurized water reactor (PWR) fuel by a dry refabrication process. This study focuses on the reactivity control method, which uses slightly enriched, depleted, or natural uranium to minimize the additional cost on the manufacturing of DUPIC fuel, when adjusting the excess reactivity of the spent PWR fuel. To reduce the variation of isotopic composition of the DUPIC fuel, interassembly mixing operations were assumed to be carried out three times. Three options have been considered: reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high-reactivity spent PWR fuels, and reactivity control by natural uranium for intermediate-reactivity spent PWR fuels. The results of this study have shown that the reactivity of DUPIC fuel can be tightly controlled with the minimum amount of fresh uranium feed. For reactivity control by slightly enriched and depleted uranium, all spent PWR fuel can be utilized as DUPIC fuel, and the fraction of fresh uranium feed is 3.4% on the average. For reactivity control by natural uranium, ~88% of spent PWR fuel can be utilized as DUPIC fuel when intermediate-reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible.

Compatibility Analysis on Existing Reactivity Devices in CANDU 6 Reactors for DUPIC Fuel Cycle

The performance of reactivity devices for a Canada deuterium uranium (CANDU) 6 reactor loaded with Direct Use of Spent Pressurized Water Reactor Fuel In CANDU reactors (DUPIC) fuel is assessed. The reactivity devices studied are the zone controller units, the adjuster rods, and the mechanical control absorbers. For the zone controller system, the bulk reactivity control, spatial power control, and damping capability for spatial oscillation are investigated. For the adjusters, the xenon override, restart after a poison-out, shim operation, and power step-back capabilities are confirmed. The mechanical control absorber is assessed for the function of compensating temperature reactivity feedback following a power reduction. This study shows that the current reactivity device system of a CANDU 6 reactor is compatible with DUPIC fuel for normal and transient operations.

Composition Adjustment on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU

In the DUPIC fuel cycle, spent pressurized water reactor (PWR) fuel is refabricated as a DUPIC fuel by a dry process. Because the spent PWR fuel composition depends on the initial enrichment and burnup conditions of PWR fuel, the composition of DUPIC fuel is not uniquely defined. To reduce the effects of such a composition heterogeneity on core performance, an adjustment of DUPIC fuel composition was studied. The composition adjustment was made in two steps: mixing two spent PWR fuel assemblies of higher and lower 239Pu contents and blending in fresh uranium with the mixed spent PWR fuels. Because the fuel and core performances depend on both the absolute amount of fissile isotopes and the ratio of major fissile isotope contents, a parametric study was performed to determine the reference compositions of 235U and 239Pu. The reference enrichments of 235U and 239Pu were determined such that the DUPIC core performance is comparable to that of a natural uranium core with high spent PWR fuel utilization and low fuel cycle cost. Under this condition, it is possible to utilize 90% of spent PWR fuels as the DUPIC fuel formula. On average, the amounts of slightly enriched and depleted uranium used for blending correspond to 8.6 and 10.6%, respectively, of the mass of candidate spent PWR fuels.

Modeling Gamma-Ray Dose Rate from a Spent Pressurized Water Reactor Fuel Assembly

Gamma-ray dose rate distribution around a pressurized water reactor spent-fuel assembly is studied using the Monte Carlo N-particle transport code (MCNP) version 4a. A detailed rod-by-rod modeling of the assembly is utilized, showing explicitly the fuel, cladding, control rod channels, and the instrumentation tube. A cylindrically distributed source of gamma rays, within every fuel rod, is considered with a seven-group energy spectrum. Dose rates are obtained by tallying the gamma rays at several axial and radial positions outside the assembly. The results indicate that the radial distribution of the dose rate can be represented by a power relationship of the form r−n, where r is the radial distance from the assembly center. Another important conclusion from this study is that the dose rate close to the assembly surface is overestimated if a homogeneous assembly model is used.

Physics Study on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU (DUPIC)

Physics studies have been performed for the compatibility of DUPIC fuel design to the existing CANDU-6. The DUPIC fuel is made of spent pressurized water reactor fuel through a dry process, and the bundle design utilizes an advanced CANDU fuel bundle geometry that is mechanically compatible with the current fuel channel and refueling system. The characteristics of a DUPIC core are compared to the current operation limits and performances of a natural uranium core. The refueling simulations have shown that the channel and bundle powers are well below the operation limits for the two- and four-bundle shift refueling schemes. The fuel performance parameters during the refueling operation reserve enough margin to the stress corrosion cracking threshold of natural uranium fuel. With the aid of burnable poison material in the fuel, the safety performance of a DUPIC core is made comparable to that of a natural uranium core. The details of DUPIC fuel design and analysis are described.

High Temperature Postirradiation Materials Performance of Spent Pressurized Water Reactor Fuel Rods under Dry Storage Conditions

Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571°C in limited air and inert gas atmospheres. N...