Burnup Values Research Articles

Reactivity and neutron emission measurements of highly burnt PWR fuel rod samples

Fuel rods with burnup values beyond 50 GWd/t are characterised by relatively large amounts of fission products and a high abundance of major and minor actinides. Of particular interest is the change in the reactivity of the fuel as a function of burnup and the capability of modern codes to predict this change. In addition, the neutron emission from burnt fuel has important implications for the design of transport and storage facilities. Measurements have been made of the reactivity effects and the neutron emission rates of highly burnt uranium oxide and mixed oxide fuel rod samples coming from a pressurised water reactor (PWR). The reactivity measurements have been made in a PWR lattice in the PROTEUS zero-energy reactor moderated in turn with: water, a water and heavy water mixture and water containing boron. A combined transport flask and sample changer was used to insert the 400 mm long burnt fuel rod segments into the reactor. Both control rod compensation and reactor period methods were used to determine the reactivities of the samples. For the range of burnup values investigated, an interesting exponential relationship has been found between the neutron emission rate and the measured reactivity.

Increased fuel burn up in a CANDU thorium reactor using weapon grade plutonium

Weapon grade plutonium is used as a booster fissile fuel material in the form of mixed ThO 2/PuO 2 fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup. Two different fuel compositions have been used: (1) 97% thoria (ThO 2) + 3%PuO 2 and (2) 92% ThO 2 + 5% UO 2 + 3% PuO 2. The latter is used to denaturize the new 233U fuel with 238U. The temporal variation of the criticality k ∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k ∞ = ∼1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the second year and remains above k ∞ > 1.06 for ∼20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up could be achieved with the same fuel material (up to 500,000 MW·D/T), provided that the fuel rod claddings would be replaced periodically (after every 50,000 or 100,000 MW·D/T). The reactor criticality will be sufficient until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

Corrections to the 148Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of 148Nd; (2) neutron capture reactions of 147Nd and 148Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of 148Nd has an effect of 1.5–2.0% on burnup values. Considering the neutron capture reactions of 147Nd and 148Nd removes dependence of C/E values of 148Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated 148Nd concentrations and the PIE data is approximately 1%, whereas this was 3–5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2–3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous analyses, does not result from a geometry model.

Investigation of CANDU reactors as a thorium burner

Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in the form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates the prospects of exploiting the rich world thorium reserves in CANDU reactors. Two different fuel compositions have been selected for investigations: (1) 96% thoria (ThO2) + 4% PuO2 and (2) 91% ThO2 + 5% UO2 + 4% PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the reactor criticality k∞ and the burn-up values of the reactor have been pursued by full power operation for >∼8 years. The reactor starts with k∞ = ∼1.39 and decreases asymptotically to values of k∞ > 1.06, which is still tolerable and useable in a CANDU reactor. The reactor criticality k∞ remains nearly constant between the 4th year and the 7th year of plant operation, and then, a slight increase is observed thereafter, along with a continuous depletion of the thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn-up can be achieved with the same fuel (>160,000 MW D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burned up, provided that the fuel rods could be fabricated to withstand such high burn-up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically.

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO 2 or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.

Effect of Neutron Induced Reactions of Neodymium-147 and 148 on Burnup Evaluation

Burnup is an important parameter in criticality safety evaluations of spent nuclear fuel in which burnup credit is taken into account. The Neodymium-148 method is widely used to evaluate the burnup of post irradiation examination (PIE) samples, and it is well known for its good accuracy. However, accuracy of the evaluated burnup values may be affected by the neutron capture reaction of 147Nd and 148Nd. Moreover, in the analysis of PIE data from a PWR, the calculation results of 148Nd have more than a 1% deviation from the experiment. In this study, the contribution of neutron capture reactions of 147Nd and 148Nd to the amount of 148Nd is discussed. The PIE data analyses using new evaluation of 147Nd capture cross section show that the JENDL-3.2 cross section data is overestimated. The change in the amount of 148Nd due to both reactions is less than 0.7% under normal reactor operation conditions. In particular, it is in the 0.1% range if burnup is approximately 30 GWd/t for a BWR and 40 GWd/t for a PWR.

The Application of U-Np Fuel and 6Li Burnable Poison for Space Reactors

The possible application of 6Li as a burnable poison and U-Np nitride as a fuel for space nuclear reactors has been studied. The analysis was performed for an infinite lattice with a leakage in the form of buckling and (U-Np)N fuel with 20% uranium enrichment. The combination of 7Li as a coolant and 6Li as a burnable poison results in a favorable criticality behavior during burnup. The parameters taken into consideration include the different fuel and coolant compositions, the form of absorber material, and the various absorber mass and concentrations. It was found that absorption properties of 6Li allow reaching the burnup value up to 67 GWd/tHM while reactivity swing is comparable with βeff. The corresponding reactor lifetime is ~10 to 30 yr.

The Ratio of Plutonium Isotopes Depending on the Burnup of the Fuel from the Fourth Block of the Chernobyl Nuclear Power Plant

The ratio of plutonium isotopes can be used for determining the burnup of nuclear fuel in the fuel-containing masses from the fourth block of the CNPP. The burnup values obtained by two independent methods using cesium and plutonium isotopes are in satisfactory agreement.

Effect of Transplutonium Doping on Approach to Long-life Core in Uranium-fueled PWR

The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them 241Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides 242Cm and 244Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile 239Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of 241 Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100GWd/tHM with about 20% 238Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles.

Considerations pertaining to the achievement of high burn-ups in HTR fuel

The satisfactory irradiation performance of coated fuel particles up to burn-ups of, say 10%, has been demonstrated in the past. However, it is shown that some of the most important physical properties of their constituents, which determine this good performance, are only poorly known. This ignorance is likely to become more serious when particles are required to attain higher burn-up values. For example, neutron dose limits above which the anisotropy of the pyrocarbon layers become unacceptably high need to be established. Again, the particle must be designed with adequate voidage to accommodate the gas released by fission, and kernel-coating mechanical interaction must be avoided. Experiments are proposed, in addition to the irradiation and post irradiation examination of particles, which should address these issues. The way this new information can be used in fuel performance codes is discussed, thereby providing a scientific understanding of the behaviour of coated particles—as of interest to researchers, by industry, by utilities and, perhaps most importantly, by regulators.

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU −1. Characteristic test of this device was done with a 252Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU −1 at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.

Long-life Water Cooled Small Reactor with U—Np—Pu Fuel

The paper presents the advanced concept of a long-life small light water reactor in which the fuel irradiation time is comparable with reactor life-time. The equilibrium analysis reveals that the U-Np-Pu fuel with unique neutronic properties allows to keep sufficient criticality up to burnup value about 140GWd/tHM. The fuel recycle does not lead to additional Pu accumulation. Both Pu and Np are well protected against un-controlled proliferation by a large fraction of 238Pu in their mixture. To improve the reactor safety, the wider fuel pin lattice was applied. The radiation damage of structural materials is within the stainless steel limitation.

The concept and neutronic characteristics of a multipurpose fast reactor (MPFR)

The paper describes a concept and its neutronic characteristics of a long-life multipurpose nuclear reactor with self-sustained liquid metallic fuel, which is proposed to meet the requirements for the energy production in the future. The application of the liquid plutonium–uranium metallic alloys as a nuclear fuel demonstrates high potential to reach excellent conversion characteristics and high burn-up values with relatively small reactivity swings. Considerations about the capability of in-vessel separation of the main fission products to prolong the core lifetime without fuel reloading or shuffling are also discussed from the viewpoint of the influence on core burn-up characteristics.

Long-life Water Cooled Small Reactor with U-Np-Pu Fuel.

The paper presents the advanced concept of a long-life small light water reactor in which the fuel irradiation time is comparable with reactor life-time. The equilibrium analysis reveals that the U-Np-Pu fuel with unique neutronic properties allows to keep sufficient criticality up to burnup value about 140GWd/tHM. The fuel recycle does not lead to additional Pu accumulation. Both Pu and Np are well protected against un-controlled proliferation by a large fraction of 238Pu in their mixture. To improve the reactor safety, the wider fuel pin lattice was applied. The radiation damage of structural materials is within the stainless steel limitation.

Temperature Reactivity Effects in Pebbles of a High-Temperature Reactor Fueled with Reactor-Grade Plutonium

This work presents the temperature reactivity effects occurring in pebbles of a high-temperature reactor fueled with reactor-grade plutonium, without any additional resonance absorbers. Burnup calculations are performed for pebbles loaded with various amounts of plutonium per pebble. During burnup, branching calculations are carried out to calculate k∞ as a function of the uniform temperature. For a high plutonium mass per pebble and low burnup values, k∞ decreases with uniform temperature, which indicates a negative uniform temperature coefficient of reactivity (UTC). However, for a low plutonium mass per pebble, as well as for a high plutonium mass per pebble in combination with high burnup, k∞ is maximal at a particular uniform temperature. Below this temperature, the UTC is positive, while above this temperature it is negative. Branching calculations with only a varying moderator temperature show almost the same behavior, which indicates that the contribution of the fuel temperature plays a minor role for the mentioned effect. To understand the reactivity effect, the moderator temperature coefficient (MTC) is investigated by two methods. In the first method, changes in reaction rates of individual nuclides and their corresponding contributions to the MTC are calculated, whereas in the second method the four-factor formula has been used. For the fresh fuel cases, the positive coefficient at low temperature is due to a positive coefficient for the thermal utilization factor. For high moderator temperatures, the coefficient for the resonance escape probability renders the MTC negative. This trend is basically caused by the shift of the high-energy tail of the Maxwell-spectrum toward the 1-eV resonance of 240Pu, and toward the resonances of 241Pu, which leads to an increase of the capture-to-fission ratio.

Gamma-ray spectroscopy on irradiated MTR fuel elements

The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

Calibration of Burnup Monitor Installed in Rokkasho Reprocessing Plant

Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid.

Neutron Intensity Measurements of BWR Spent Fuels

A neutron scanning device was developed in order to obtain accurate neutron intensities of high burn-up BWR fuels. This scanning device was calibrated with a 252Cf source and used to measure axial distributions of neutron intensities of BWR fuels with various enrichments (2.0%–3.4%) irradiated up to 60 GWd/tU at Fukushima Daini Nuclear Power Station Unit 2(2F-2). The measured neutron intensities were approximated well with power law interpolations on the calculated burn-up values. The neutron intensities calculated by the ORIGEN2-86 code showed good agreements with the measured ones within 20%.

An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/ 97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/ 137Cs ratio and 106Ru/ 137Cs ratio.

Advanced U–Np–Pu fuel to achieve long-life core in heavy water reactor

The objective of this paper is to look at the possibility of approaching the long-life core comparable with reactor life-time. The main issues are centered on U–Np–Pu fuel in a tight lattice design with heavy water as a coolant. It is found that in a hard neutron spectrum thus obtained, a large fraction of 238Pu produced by neutron capture in 237Np not only protects plutonium against uncontrolled proliferation, but substantially contributes in keeping criticality due to improved fissile properties (its capture-to-fission ratio drops below unit). Equilibrium fuel composition demonstrates excellent conversion properties that yield the burn-up value as high as 200 GWd/t at extremely small reactivity swings.