Burnup Determination of High Burnup and Dry Processed Fuels Based on Isotope Dilution Mass Spectrometric Measurements

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, ¹⁴⁸Nd, ¹⁴⁵Nd+¹⁴⁶Nd, total of the Nd isotopes, ¹³³Cs and ¹³⁷Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes (²³³U, ²⁴²Pu, ¹⁵⁰Nd, and ¹³³Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.     

日本六个所后处理厂减少废物产生量的工艺设计

为减少放射性废物的产生量,日本六个所后处理厂在建造过程中吸收了乏燃料后处理技术发展的最新成果。对原设计进行了改进,降低了废物产生量,减少了高放废液贮存罐的数目。日本核燃料有限公司和三菱重工长崎研发中心等提出了Nox回收利用工艺,也大大减少了低放射性硝酸钠废物和非放射性硝酸钠废物的产生景。本文叙述了日本六个所后处理厂减少放射性废物产生量方面的情况。     

乏燃料后处理厂主工艺中的主要化学安全问题

水法后处理工艺过程涉及很多化学反应,反应条件和反应产物不同,需要关注的化学安全问题也不同。描述了后处理主工艺不同阶段的化学反应,分析了各阶段应关注的主要化学安全问题,为商用后处理厂的设计和事故分析提供参考。     

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

Measurement of Burnup in FBR MOX Fuel Irradiated to High Burnup

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system.

Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector.     

Development of Adsorption Process for NOx Recycling in a Reprocessing Plant

A large amount of NOx, which is used in a reprocessing plant mainly as an oxidizing agent of Pu³⁺, eventually results in the formation of low-level radioactive sodium nitrate waste. Since NOx is generated by the reaction of sodium nitrite and nitric acid, non-radioactive sodium nitrate is also formed as a by-product. In order to reduce the amounts of radioactive and non-radioactive sodium nitrate wastes, a new method was examined to recover NOx for recycling from the off-gas of the denitrator of uranyl nitrate solution. Fundamental and consequent bench scale experiments showed that the vacuum pressure swing adsorption method, using combined silica-gel and clinoptilolite for water vapor removal and pentasil zeolite for NOx recovery, is applicable for this purpose.     

Environmental Tritium in the Vicinity of Tokai Reprocessing Plant

The distribution of ³H in the atmosphere, vegetations and soil water was observed in the vicinity of Tokai reprocessing plant (TRP) over the period 1990–2004. The annual means of the atmospheric HTO and HT concentrations were in the range of 12–40 mBqm^?3 with a significant seasonal variation and 14–51 mBqm^?3 with no seasonal variation, respectively. Long-term atmospheric dilution factors, defined as the annual mean of the atmospheric HTO concentration divided by the discharge rate of HTO from the TRP, were estimated to be 10^?8–10^?6 sm^?3. The atmospheric HTO concentrations decreased with distance from the TRP, falling to the current background level in Japan at 5 km off-site. The HTO concentrations observed were compared with those calculated by a simple mathematical model with input data of the monthly ³H discharge rates and actual meteorological conditions. The calculations were correlated well with the observations even for only a little HTO-elevated situation, considering the naturally occurring ³H level in atmospheric vapor around the TRP. Tritium concentrations in vegetation and soil water samples were roughly the same as the atmospheric HTO concentrations, suggesting the rapid equilibrium of ³H concentrations in the atmosphere-soil-vegetation system around the TRP.     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

A Large Area Thin Nal(Tl) Detector for External Counting of Low-Energy Emitters in Lung

A new type lung monitor was designed for the detection of ²³⁹Pu and ²³⁵U in the lungs, and a preliminary calibration undertaken thereon. The detector consists of a 9.5 in. dia. by 0.5 in. thick NaI(Tl)-crystal and seven photo-tubes. Besides its use as an ordinary lung monitor, the instrument can be employed as an image detector giving a rough indication of the distribution of the activities in the lungs. This paper describes the structural arrangement of the detector, the preliminary calibration performed thereon and the limits of detection derived as function of the subject's effective tissue thickness.     

Formation of Pellet-Cladding Bonding Layer in High Burnup BWR Fuels

Formation process of the pellet-cladding bonding layer was studied by EPMA, XRD, and SEM/TEM for the oxide layer on a cladding inner surface and the bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27 and 42 GWd/t in BWRs. In the lower burnup specimens of 15 and 27GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and previously reported 49 GWd/t had a typical bonding layer. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the cladding was made up mainly of ZrO₂. The structure of this ZrO₂ consisted of cubic phase, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U, Zr)O₂ and amorphous phase. Even in the lower burnup specimens having no bonding layer, cubic ZrO₂ phase was identified in the cladding inner oxide layer. The formation process of the bonding layer were discussed in connection with phase transformation by irradiation damage of fission products and conditions for contact of pellet and cladding.     

Long Term Change in Level-Volume Relation of an Accountability Tank for Plutonium Nitrate Solution

A slight change in the level-volume relation for an accountability tank for a large amount of plutonium nitrate solution (PuN) was observed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC), Tokai Works. From the results of annual tank re-calibrations for the plutonium receiving tank from 1985 to 1992 using the incremental feed of nitric acid as the density standard, it became clear that the relation between the level and the volume changed slightly, and the rate of the change was a linear function of operating time. Also it became clear that the change was linear in relation to the level. In the PCDF, the cumulative change in the volume at the nominal level was evaluated to be 0.1% during 8 years' operation. It was also evaluated that the repeatability of the re-calibration is much better than 0.1%. A reasonable frequency of tank re-calibration is once every 5 years.     

Investigation on Space Dependence of C/E Values for Control Rod Worths and Reaction Rate Distributions in ZPPR-10A and -10D Assemblies

In order to investigate the redox equilibrium of uranium ions in molten NaCL-2CsCL, UV-Vis absorption spectrophotometry measurements were performed for U⁴⁺ and U³⁺ in molten NaCL-2CsCL at 923 K under simultaneous electrolytic control of their ratio. Prominent absorption bands at 480 and 570 nm were assigned to U³⁺, and their molar absorptivities were determined to be 1,260±42 and 963±32 mol^?1.l.cm^?1 respectively. From the dependence of the rest potential of the melt on the spectrophotometrically determined ratio of [U⁴⁺]/[U³⁺], the standard redox potential of the couple U⁴⁺/U³⁺ at 923K was determined to be ?1.481±0.004 V vs. Cl₂/Cl^?. Cyclic voltammetry measurements were carried out for the couple U⁴⁺/U³⁺, and the results agreed well with this standard redox potential value. By the results of cyclic voltammetry, a temperature dependence of the standard redox potential was found to be ?2.094+6.639×10^?4 T (T=823-923 K).