Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.     

Core concept of compound process fuel cycle

This paper presents fast reactor core concept and its feasibility as a part of newly proposed compound process fuel cycle in which spent fuels of light water reactor are multi-recycled without conventional reprocessing but with only pyrochemical processing, fuel re-fabrication and reloading to the fast reactor core. Results of the core survey analyses in order to find out the feasibility of this concept, taking example for BWR MOX spent fuel of 60 GWd/t burn-up, show that four times recycling of LWR spent fuel with the burn-up of more than 300 GWd/t can be achieved without increasing MA content. Such benefits will be expected in this concept as reduction of fuel cycle cost due to simplified reprocessing procedure, reduction of environmental impacts due to reduced high level waste, efficient utilization of nuclear fuel resources, enhancement of nuclear non-proliferation, and suppression of LWR spent fuel pile-up.     

Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 × 8 BWR fuel assembly are low up to 60 GWd/t.     

Oxygen stoichiometry shift of irradiated LWR-fuels at high burn-ups: Review of data and alternative interpretation of recently published results

The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release.     

Post-irradiation examination of uranium-doped rock-like oxide fuels

Post-irradiation examinations of rock-like oxide fuels were performed in JAERI to evaluate irradiation behavior and geochemical stability. Five kinds of fuels were prepared using 20% enriched U instead of Pu; a single-phase fuel of an yttria-stabilized zirconia containing UO₂ (U-YSZ), two particle-dispersed type fuels of U-YSZ particles + MgAl₂O₄/Al₂O₃ powder, two homogeneously blended type fuels of U-YSZ powder + MgAl₂O₄/Al₂O₃ powder. The fuels were irradiated in JRR-3 for about 100 days and estimated irradiation conditions were as follows; linear power was 15 kW m⁻¹, thermal neutron fluence was 7 x 10²⁴ m⁻² and fuel temperatures at the surface were 740–1130 K. From the results of non-destructive examinations, the stainless steel cladding surfaces were partially discolored by oxidation and no remarkable deformation of the pins was observed. Significant pellet fragmentation was not observed in spite of the crack formation as observed in irradiated LWR UO₂ fuels. Nonvolatile FPs were observed only in pellets but volatile Cs moved partly to the plenum. From these examinations, no significant difference in macroscopic irradiation behavior was distinguished among 5 fuels.     

Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead–bismuth, Pb–Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

Recycle transmutation of MA and LLFP using BWR for sustaining geologic disposal

The actinides and fission products produced in nuclear fuels constitute an important part of the HLW. Therefore, methods for reducing the radiotoxicity of the MA and LLFP in HLW are presently under investigation. The purposes of this study are to evaluate the effectiveness of MA transmutation by taking advantage of neutron spectrum hardening due to void fraction along BWR axial direction; to understand the effectiveness of LLFP transmutation in BWR considering the large capture cross section of FP in thermal region; and to evaluate the macroscopic characteristics of longer residential period of LLFP target in the high burnup BWR core. Conceptual B/T BWR supposed in this study was reactor which the performance comparable to the current BWR. In MA transmutation case, the calculation was focused on varying the void fraction of 0 to 40% along the axial direction, which were directly associated to the lower and upper region of the BWR core. The performance of B/T BWR was evaluated in which four components of MA (²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm) with fixed fraction were blended with UO₂ in B/T fuel. While, for LLFP transmutation, the B/T BWR was assumed to have two homogeneous regions: {1} the region for UO₂ driver fuel (99% of fuel weight), and {2} the region for LLFP (⁹⁹Tc and ¹²⁹I) target capsules (1% of fuel weight), in which metallic Tc rods and iodine in the form of CeI₃ was contained in cylindrical target capsules. The evaluation functions are {1} fission-to-transmutation ratio, [F/T ratio]_MA, and {2} transmutation fraction, Tf_LLFP. Results show that the hardening neutron spectrum due to increase of void fraction in B/T BWR would result a higher [F/T ratio] of MA transmutation performance. Np and Am would be effectively loaded in the upper region of the core, while Cm could be loaded in any region of the core. At the EOC of equal or more than 50 GWd/Mg(HM), technetium has a higher transmutation fraction compared to iodine. To obtain higher LLFP transmutation fraction, the residential time in the LLFP targets in the core, should be kept for long time, for instance about 10 to 30 years. For that purpose, it was proposed that the number of B/T BWR system for LLFP treatment corresponds to the residential time of the LLFP target, i.e. 10 to 30 units.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Neutron beam design for both medical neutron capture therapy and industrial neutron radiography for TRIGA reactor

Neutron beam designs were studied for TRIGA reactor with a view to generating thermal, epithermal and fast neutron beams for both medical neutron capture therapy (NCT) and industrial neutron radiography (NR). The beams are delivered from thermal and thermalizing columns, and also horizontal beam hole. Several prospective neutron filters (high-density graphite (G), bismuth (Bi), single-crystal silicon (Si), aluminum (Al), aluminum oxide (Al₂O₃), aluminum fluoride (AlF₃) and lead fluoride (PbF₂)) were examined for obtaining sufficiently intense neutron beam for various applications. Monte Carlo calculations indicated that with a suitable neutron filter arrangement, thermal and epithermal neutron beams attaining 2×10⁹ and 7×10⁸ n cm⁻²S⁻¹, respectively, could be obtainable from thermal and thermalizing columns with the reactor operating at 100 kW. These neutron beams could be adopted for boron neutron capture therapy. Compared with these columns, horizontal beam port would deliver neutron fluxes of 10⁻² 10⁻³ lower intensity, but produced thermal and neutron beams would be adequate for different application of nondestructive inspection by neutron radiography.     

Evaluation on chemical state of irradiated rock-like oxide fuels by nuclear and chemical modeling

Chemical forms of fission products in irradiated ROX fuels were calculated by the SOLGASMX-PV code, and the resultant phase equilibrium and the oxygen potential in the fuel were evaluated in order to assess the irradiation behavior of the ROX fuels. For the ROX fuel with reactor grade Pu, the oxygen potential increased to about −140 kJ mol⁻¹ at EOL when all the Pu in the fresh fuel was tetravalent. In the case of fresh fuel which was partially reduced with the [Pu⁺³]/[Pu⁺⁴]=10/90, the oxygen potential increase was suppressed to about −400 kJ mol⁻¹. On the other hand, the oxygen potential of the ROX fuel with weapon grade Pu never exceeded the value of about −400 kJ mol⁻¹. The difference of oxygen potentials was caused by difference of Am amount produced by Pu conversion. The oxygen potential of the irradiated fuel was controlled by the phase equilibria among FPs. The equilibrium between metallic Mo and MoO₂ controlled the oxygen potential to about −400 kJ mol⁻¹.     

Lattice location and electrical conductivity in ion implanted TiO2 single crystals

Single crystals of TiO₂ (rutile) were implanted at room temperature with Ar, Sn and W ions applying fluences of 10¹⁵/cm² to 10¹⁶/cm² at 300 keV. The lattice location, together with ion range and damage distribution was measured with Rutherford Backscattering and Channeling (RBS-C). The conductivity, σ, was measured as a function of temperature. The implanted Sn and W atoms were entirely substitutional on Ti sites in the applied fluence region, where the radiation damage did not yet reach the random level. A large σ increase was observed for all implants at displacement per atom values (dpa) below 1. Above dpa = 1, σ reveals a saturation value of 0.3 Ω⁻¹ cm⁻¹ for Ar implants, while for W and Sn implants a further increase of σ up to 30 Ω⁻¹ cm⁻¹ was measured. Between 70 K and 293 K ln σ was proportional to T^−1/2, (Ar,W) and T^−1/4 (Sn), indicating that the transport mechanism is due to variable range hopping.     

Noise analysis performed at the sodium cooled power reactor KNK I for malfunction diagnosis and plant surveillance

In order to use neutron noise analysis as an effective tool for early malfunction detection it is necessary to identify the driving forces and to calculate their contributions to the power fluctuations. In this paper the influence of a considerable number of measured noise sources on neutron noise within a large frequency range (10⁻³ Hz to 10³ Hz) is investigated for the sodium cooled power reactor KNK I (thermal core, 58 MW_th).

The experimental basis for the analysis is numerous records of the following signals at various power levels: neutron noise which has been measured with an in-core fission chamber and 3 ex-core ionisation chambers; the sodium inlet temperature and the coolant flow in both primary coolant loops and the movement of the control rods. In addition signals from acoustic-, seismic- and pressure transducers and the coolant outlet temperature were collected.

The influence of the thermohydraulic- and of the control system on neutron noise has also been calculated by means of the relations for linear and multiple-input systems. Important for this analysis is the reactivity-power transfer function. Calculations of this function could be confirmed by measurements using a pseudo-random binary signal as reactivity input.

The following results were obtained from the analysis of the auto-power spectral densities of the neutron flux: Fluctuations of the coolant inlet temperature and the coolant flow are relatively small sources for neutron noise. However, reactivity adjustments resulting from the automatic control system because of the inherent instability of the reactor turned out to be an important driving force.

The influence of still unknown driving forces increased considerably with the reactor power. Since the coolant flow was proportional to the reactor power in order to keep the coolant temperature constant, this result indicates that turbulent flow must have induced stochastical movements of core components. These movements are considered to have mainly caused the unknown reactivity driving forces. Their magnitude could be determined reliably only in the frequency range, in which external feedback mechanisms through the primary coolant system were negligible. For 30 to 50 % reactor power the contribution was about 30 % (for f > 5·10⁻³ Hz) and for full power it increased to about 80 % (for f > 5·10⁻² Hz) of the measured neutron noise. For frequencies > 5 Hz the white detection noise prevails. Single peaks in this frequency region could be explained by coherence function investigations between in-core and ex-core neutron detector signals and by correlation of these signals with displacement- and pressure fluctuations.

Though the measured neutron noise could not be unambiguously related to driving forces, the combination of analytical and empirical methods makes the results also applicable for the design of surveillance techniques for other sodium cooled reactors (e.g. LMFBRs). Examples for possible applications are given.     

利用测热技术测量核反应堆中子通量密度

一种新型中子探测器被研究,其原理是利用带电离子在矿物中沉积的能量退火时会以热量的方式释放出来,通过测量释放的热量而确定中子通量密度。对新型中子探测器进行刻度,在反应堆内某位置测量的热中子通量密度为5.108×10¹¹ cm^-2•s^-1,与标定的热中子通量密度(5.000×10¹¹ cm^-2•s^-1)在2%内符合,说明该探测器可测量中子通量密度。本文方法制作的探测器体积小,可制作成不同形状,便于反应堆不同环境下的中子通量密度测量。选取相应中子能量反应截面较大的元素,该探测器还可测量不同中子能量的通量密度。     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Investigation of in-reactor creep and irradiation growth of zirconium-2.5 wt% niobium channel tubes

We summarize the diametral creep results obtained in the MR reactor of the Kurchatov Institute of Atomic Energy on zirconium-2.5 wt% niobium pressure tubes of the type used in RBMK-1000 power reactors. The experiments that lasted up to 30 000 h cover a temperature range of 270 to 350°C, neutron fluxes between 0.6 and 4.0 ×10¹³ n/cm² · s (E > 1 MeV) and stresses of up to 16 kgf/mm². Diametral strains of up to 4.8% have been measured. In-reactor creep results have been analyzed in terms of thermal and irradiation creep components assuming them to be additive. The thermal creep rate is given by a relationship of the type ε_th = A₁ exp [(A₂ + A _3σt) T] and the irradiation component by ε_rad = A_4σtø(T − A₅), where T = temperature, σ_t = hoop stress, ø = neutron flux and a₁ to A₅ are constants. Irradiation growth experiments carried out at 280° C on specimens machined from pressure tubes showed a non-linear dependence of growth strain on neutron fluence up to neutron fluences of 5 × 10²⁰ n/cm². The significance of these results to the elongation of RBMK reactor pressure tubes is discussed.     

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Irradiation growth during fission fragment and proton bombardment of uranium

Irradiation growth results are reported for annealed -uranium at 373 K under 3.5 MeV proton bombardment. Two such experiments were performed at damage rates of 6.9 × 10⁻⁸ and 9.3 × 10⁻⁸ dpa/s to doses of 0.0072 and 0.0077 dpa, respectively. In each case the growth rate remained constant throughout the experiment. The respective damage normalised growth rates were 5.6 × 10⁻³ and 7.1 × 10⁻³ dpa⁻¹. Comparison between proton growth rates and published in-reactor growth rates is made by converting the more usual fuel damage parameters, such as burn-up, to dpa. Damage calculations, using the NRT damage model, are presented which indicate that, in uranium, each fission event produces 100 000 displacements. The reported growth rate of annealed, polycrystalline -uranium at 353 K, during thermal neutron irradiation, represents a damage normalised growth rate of 9.6 × 10⁻³ dpa⁻¹, which is not substantially different from the present proton results. This similarity of proton and fission growth rates appears to be contrary to the earlier finding of Thompson (1960), who deduced that proton bombardment produced two orders of magnitude less growth than fission fragments. Thompson concluded that thermal spikes played a dominant role in irradiation growth. Thompson's results and analysis are reassessed in the light of recent range data and damage models and found to be consistent with the present results in both magnitude and direction. The results are also inconsistent with Buckley's original model to the extent that thermal spikes were thought to play an important role. From a consideration of primary recoil spectra it is shown that the concept of the anisotropic aggregation of point defects to form vacancy and interstitial clusters, which is at the centre of that model, remains viable. Furthermore, similar though slightly less growth would be expected during proton bombardment. This was indeed found to be the case, the growth rate with protons being about half that with fission fragments.     

医院中子照射器中子束流出口处热中子注量率的测量

医院中子照射器是基于微型反应堆而设计的专门用于硼中子俘获治疗（BNCT）的核反应堆装置,其额定功率为30 kW。在堆芯相对两侧分别设有一条热中子束流和超热中子束流用于病人照射,在热中子束流内引出一条实验用热中子束流,用于瞬发γ法测量病人血硼浓度。本工作利用²³⁵U裂变靶和白云母探测片测量了热、超热和实验用热中子束流出口处的热中子绝对注量率。结果显示,在30 kW额定功率运行时,热、超热和实验用热中子束流出口处的热中子注量率分别为1.67×10⁹、2.44×10⁷和3.03×10⁶ cm^-2•s^-1。以上结果达到了BNCT设计要求,并能满足瞬发γ测量血硼浓度的要求。