Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are ① low enriched uranium (LEU); ② weapon grade plutonium + ThO₂; ③ reactor grade plutonium + ThO₂; and ④ minor actinides in the spent fuel of light water reactors (LWRs) + ThO₂. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs.The criticality calculations are conducted with SCALE5.1 using S₈–P₃ approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium.Using LEU, an enrichment grade of 9% has resulted with k_eff = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO₂ + 80% ThO₂ yields k_eff = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO₂ + 65% ThO₂ brings it to k_eff = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k_eff = 1.2673 is achieved by 50% MAO₂ + 50% ThO₂.The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.     

Criticality aspects of nuclear power reactor cores in the case of emerging nuclear fuels

Reactor cores of PWR and LMFBR, loaded with different commercial and emerging nuclear fuels, have been simulated and compared at BOI with respect to criticality with and without chemical shim, control rods and sodium. The different cases considered, within each of the reactor types, are grouped together according to their fissile content, when compared on the basis of the neutron multiplication factor (k_eff). For both PWR and LMFBR reactor types, the reactivity worths of the control rods do not change significantly when replacing commercial fuels by emerging ones. In the case of the LMFBR, the Na void reactivity effects are small and comparable using either emerging or commercial fuels. Hence, operation and control of the core at beginning of irradiation are similar for emerging or commercial fuels.     

Criticality qualification of a new Monte Carlo code for reactor core analysis

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.     

Experimental investigation of fluctuations in a pulsed reactor

The paper is devoted to the experimental investigation of fluctuations in intensity of power excursions in the IBR pulsed reactor in the Joint Institute of Nuclear Research. The measurements were made, using a scintillation detector, at six reactor power levels from 30 to 1200 hr. The relationships obtained can be recommended for studying problems concerning the physics of reactors of this type.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 12–16, January, 1964     

Critical fluctuations of the number of neutrons in a reactor

Translated from Atomnaya Energiya, Vol. 78, No. 6, pp. 366–376, June, 1995.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Criticality calculations of spent fuel in deep geological repository

Criticality calculations have been performed for a typical spent fuel disposal canister model filled with PWR fuel elements. Geometric and material properties of the disposal canister and disposal holes were modeled based on the Swedish preliminary disposal concept. Direct disposal of 5% enriched 16 × 16 PWR fuel was considered. We performed the calculations of the neutron multiplication factor using various disposal configurations, depending on the initial enrichment, fuel burnup, canister geometry and disposal holes configuration. The results showed that under normal conditions, when the canister is filled with fresh spent nuclear fuel, the system is deeply sub-critical. If it is assumed that the canister is faulty, leaking and filled with ground water, the system may become critical in the case of fresh fuel.     

Criticality calculation for cluster fuel bundles using grey Dancoff factor

This paper applies the grey Dancoff factor calculated by Monte Carlo method to the criticality calculation for cluster fuel bundles. Dancoff factors for five symmetrically different pin positions of CANDU37 and CANFLEX fuel bundles in full three-dimensional geometry are calculated by Monte Carlo method. The concept of equivalent Dancoff factor is introduced to use the grey Dancoff factor in the resonance calculation based on equivalence theorem. The equivalent Dancoff factor which is based on the realistic model produces an exact fuel collision probability and can be used in the resonance calculation just as the black Dancoff factor. The infinite multiplication factors based on the black Dancoff factors calculated by collision probability or Monte Carlo method are overestimated by about 2 mk for normal condition and 4 mk for void condition of CANDU37 and CANFLEX fuel bundles in comparison with those based on the equivalent Dancoff factors.     

Criticality risks during transportation of spent nuclear fuel

Abstract

A probabilistic risk assessment (PRA) quantifies the frequency of criticality accidents during railroad transport of spent nuclear fuel casks (SFCs) in the USA. It evaluates the likelihood that undetected errors in fuel selection and/or fuel handling could result in a misloaded SFC susceptible to a criticality event following an accident during rail transport of the cask. The PRA shows that existing fuel burnup records and formal procedures for loading a SFC make the likelihood of shipping a misloaded SFC on the order of 2·6 × 10^–6 per SFC. When combined with historical evidence regarding train accidents and an estimate of the likelihood that an accident could breach and submerge a SFC, the calculated frequency of criticality is below 2 × 10^–12 over the 11 000 shipments that would be required to ship the spent fuel inventory generated by the current US fleet of nuclear reactors, assuming that they each operate for 60 years.     

Power flattening in the fuel bundle of a CANDU reactor

The strong non-uniformity of the fission power production density in the CANDU fuel bundle could have been mitigated to a great degree. A satisfactory power flattening has been achieved through an appropriately evaluated method by varying the composition of the LWR spent fuel/ThO₂ mixture in a CANDU fuel bundle in radial direction and keeping fuel rod dimensions unchanged. This will help also to greatly simplify fuel rod fabrication and allow a higher degree of quality assurance standardization.Three different bundle fuel charges are investigated: (1) the reference case, uniformly fueled with natural UO₂, (2) a bundle uniformly fueled with LWR spent fuel, and (3) a bundle fueled with variable mixed fuel composition in radial direction leading to a flat power profile (100% LWR spent fuel in the central rod, 80% LWR + 20% ThO₂ in the second row, 60% LWR + 40% ThO₂ in the third row and finally 40% LWR + 60% ThO₂ in the peripheral fourth row).Burn-up grades for these three different bundle types are calculated as 7700, 27,300, and 10,000 MW.D/MT until reaching a lowest bundle criticality limit of k_∞ = 1.06. The corresponding plant operation periods are 170, 660, and 240 days, respectively.