G. HORDÓSY
KFKI Atomic Energy Research Institute, Budapest, Hungary
Email: hordosy@sunserve.kfki.hu
Abstract
According to the present Hungarian regulations, the criticality safety analysis of a transport/storage device should be based on the fresh fuel assumption. It was recognized that for new advanced fuel types with higher enrichment, some of the existing transport/
storage facilities can only be used with reduced capacity unless burnup credit is used in the subcriticality analysis. The impact of burnup credit on the capacity of these facilities, for the compact storage pool and for the TK-6 transport cask is examined. Preliminary investigations have been made on some aspects of the safe application of burnup credit.
These are the study of the influence of uncertainty in nuclear data on criticality calcula-tions, the study of the influence of the axial burnup distribution on criticality and the testing of the depletion calculation methodology with newly available post-irradiation measurements.
1. INTRODUCTION
Hungary has four WWER-440 units at the Paks nuclear power plant.
These four units provided 38.6% of the total production of electric energy in the country in 2005. The original fuel assembly has a 1.22 cm pitch and 3.6%
maximum enrichment. Recently the enrichment was increased to 3.82% with radial zoning, and a change of lattice pitch from 1.22 cm to 1.23 cm is now being considered.
After removal from the core the spent fuel is loaded into the wet storage pool at the reactor. The pool contains two racks. The upper rack has a 22.5 cm spacing between the assemblies; subcriticality is ensured by the water among the assemblies. The lower rack has compact storage; it contains boron steel plates and the assembly spacing is 16 cm. After some years of cooling the spent assemblies are transported either to the dry interim storage module using a C-30 transport cask or to the Russian Federation using a TK-6 transport cask.
The maximum storage capacity of both casks is 30 assemblies.
According to Hungarian regulations, the criticality safety analysis of a transport/storage device must be based on the fresh fuel assumption. The effective multiplication factor should be less than 0.95 for all normal conditions and should be less than 0.98 for any single failure case. Meeting these subcriti-cality requirements should be ensured by a conservative safety margin covering all kinds of uncertainties. This requirement can be written as:
kc + Δkc ≤ USL (1)
where kc is the calculated value of the multiplication factor, Δkc is the error of the calculation and USL is the upper safety limit, which already contains the administrative safety margin (usually 0.5) and all of the uncertainties, including the error in the estimation of cross-sections. This uncertainty should be determined using critical benchmark experiments on assemblies similar to the device being investigated. For fresh fuel it can be derived from UO2 criticality experiments.
If the full capacity is used, compliance with the subcriticality criteria can be demonstrated by the fresh fuel assumption for the upper rack of the storage pool, for the interim dry storage modules and for the cask used for transport between the pool and the dry storage module. However, using the fresh fuel assumption, the 3.82% fuel in the compact rack and 3.6% fuel in the TK-6 cask under optimal moderation conditions do not meet the subcriticality criteria if they are used to full capacity. At present, subcriticality is ensured by technical measures in these facilities, i.e. they are used with a reduced capacity.
Preliminary investigations of the benefits of using burnup credit for these facilities were carried out. They are described in Sections 2 and 3. For safe application of burnup credit, the uncertainties associated with the cross-section and with the accuracy of composition calculations should be analysed. The possibility of performing these investigations using the available experimental data is investigated in Sections 4 and 5.
2. SUBCRITICALITY OF THE COMPACT STORAGE RACK
The compact storage pool has its maximum multiplication properties with a water density of 1 g/cm3. In the case of assemblies with 3.82% enrichment and 1.23 cm pitch the subcriticality criteria are met if absorber assemblies are placed in every 25th position. This means a reduction of approximately 4% in storage capacity. This capacity decrease could be avoided by application of burnup credit.
POTENTIAL BENEFITS OF BURNUP CREDIT IN HUNGARY
For reliable determination of the required average assembly burnup, the influence of the burnup distribution on the multiplication factor should be taken into account. From previous experience it is known that keff is not determined unequivocally by the average assembly burnup, but may depend significantly on the axial change of the burnup. This influence is generally referred to as the ‘end effect’ because of the importance of the top and bottom end of the fuel assembly having lower burnup. This phenomenon may vary from plant to plant, depending on the fuel type, the loading strategies, etc.
To account for this influence in this particular case a number of criticality calcu-lations were performed for the compact storage rack, using real axial assembly burnup distributions as well as with uniform burnup distributions corre-sponding to the average of the real axial burnup distribution. The axial burnup distributions were ‘real life’ distributions, derived from the KOLA benchmark [1]. This benchmark definition contains the reload patterns and detailed operational histories of the first 12 cycles of unit 3 of the KOLA nuclear power plant. From the fifth cycle the core contained assemblies with 4.4%
enrichment. Several assemblies achieved average burnup levels up to 50 MW·d/kg U and a few assemblies even higher levels. The reason for choosing this benchmark as a source for the distribution is that higher enrichment fuel (4.4%) was used and higher burnup was achieved than is the case for the usual WWER-440 unit. Details of the burnup and composition calculations are given in Ref. [2]. The calculations were performed using
‘actinides only’ and ‘actinides + fission products’ burnup credit. The actinides U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243 and the fission products Mo-95, Tc-99, Ru-101, Rh-103, Ag-109, Cs-133, Nd-143, Nd-145, Sm-147, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Gd-155 were considered. About 200 axial assembly burnup distributions were investigated. Assemblies with 3.6% and 4.4% enrichment were considered.
Approximately the same maximum end effect was found for 3.6% and 4.4% assemblies. This value was about 3% using fission products + actinides and about 1.5% using actinides only. The influence of the radial burnup distri-bution was also investigated for 20 assemblies which were close to the core boundary during their last cycle in the core. No statistically significant influence on keff was found.
Taking into account the effect of the axial distribution it was found that the subcriticality of the compact rack can be ensured if the assembly burnup B meets the requirement that
B > 11 MW·d/kg U using the actinides only approach, and
B > 8.5 MW·d/kg U if the actinides + fission products approach is used.
It is noted that these values of burnup are generally achieved during one cycle.
3. SUBCRITICALITY OF THE TK-6 TRANSPORT CASK
The TK-6 transport cask is intended to be used for transporting spent fuel assemblies to the Russian Federation. It is a steel container 392 cm in height and with an outer radius of 106 cm. Inside the container there is a basket containing the assemblies in a hexagonal lattice. The lattice pitch of the assembly is 22.5 cm. A maximum of 30 assemblies can be loaded into the cask.
During the transport operation, the assemblies are stored under water. Above the water level the cask is filled with nitrogen. The horizontal cross-section of the cask is shown in Fig. 1.
For normal water density this cask is subcritical with 30 assemblies of the type described above. However, there is a water drain valve at the bottom of the cask. An accident that involves the valve being broken cannot be excluded.
In this case the water would leak out at the bottom of the closed cask and the water level and the pressure above the water level would decrease. A thermo-hydraulic analysis of this situation is not available, so an investigation of the optimal moderation conditions is necessary. The results of the analysis show that a cask containing 30 fuel assemblies with 3.6% enrichment would not meet the subcriticality criteria. Subcriticality can be ennsured by technical measures,
FIG. 1. Horizontal cross-section of the TK-6 transport cask.
POTENTIAL BENEFITS OF BURNUP CREDIT IN HUNGARY
such as loading only 24 assemblies into the cask or loading 27 assemblies and 3 absorber assemblies. This is a substantial reduction of the transport capacity, which could be avoided by using burnup credit.
Simplified analysis shows that for an assumption of about 28 MW·d/kg U average assembly burnup the cask would meet the subcriticality criteria in the optimal moderation conditions. In this analysis, only the actinides were used and the same extent of end effect was assumed as in the previous case. The calculated multiplication factor of the cask as a function of water density is shown in Fig. 2 with the fresh fuel assumption and with burnup credit.
4. INFLUENCE OF THE PLUTONIUM
AND URANIUM CROSS-SECTION UNCERTAINTIES IN BURNUP CREDIT APPLICATION
Errors in cross-section data represent a major source of uncertainty in criticality calculations. This uncertainty should be taken into account using criticality benchmark experiments. A set of criticality experiments should be selected with relevant neutron physical characteristics (enrichment, lattice pitch, hydrogen/uranium ratio, etc.) similar to the investigated application.
Criticality calculations should be performed for this set of experiments using the computational tool to be validated (code + data library). Using statistical methods for comparison of the calculated and measured multiplication factors, FIG. 2. Multiplication factor of TK-6 as a function of water density with the fresh fuel assumption and with burnup credit.
an upper safety limit (USL) containing the bias and uncertainties associated with that particular computational tool can be derived.
Such a method for the derivation of the USL was developed at the Oak Ridge National Laboratory (ORNL) in the USA. The USLSTAT code for implementing these methods was also developed at ORNL. They are described in detail in Ref. [1]. This method, involving the use of a confidence band with an administrative margin approach, has been used for criticality safety analysis in KFKI AERI in recent years. In its general form the derived USL can be written as
USL(x) = 1 – Δkm – W – β(x)
where x is the physical parameter which gives the most conservative USL in the trending analysis, Δkm is the administrative safety margin (usually 0.05), β(x) is the bias, i.e. the difference between the linear fit and the measured keff values.
W is the confidence bandwidth for the lower confidence limit. For a specified confidence level, W is determined by the deviation of calculated and measured keff values.
While it is relatively easy to find appropriate benchmark experiments for fresh fuel analysis, it is more difficult for burnup credit applications. There are plenty of UO2 critical experiments, but there are no publicly available experiments for burned fuel. For criticality, the most important elements are uranium and plutonium. A possible approach therefore is to use MOX criticality experiments for this purpose. However, the question of the extent to which these experiments are adequate for real transport/storage application may arise.
In an attempt to derive an upper subcriticality limit for burnup credit applications, 132 MOX experiments were selected and investigated using the ICSBEP Handbook. Although this is a large number of experiments, their characteristics only partly cover the neutronic features of the applications of interest (compact storage and TK-6 transport cask). The PuO2 content of the MOX fuel varies from 1.5% up to 20% in these experiments. The lower bound of this interval is close to the plutonium content of the burned fuel of interest in the case being investigated. The ‘plutonium enrichment’, i.e. the ratio of fissionable plutonium atoms to the total number of plutonium atoms, ranges from approximately 70% to 90%, which is applicable for the cases being inves-tigated, where this quantity is about 75–77%. The lattice pitch is larger in all experiments compared to those which exist in WWER-440 fuel, but this could be extrapolated. However, in all experiments, natural or depleted uranium dioxide was used, i.e. the U-235 content is only approximately 0.7 or 0.2%.