Conceptual Study of Neutron Physics of Nuclear Fuel Cycle for Ceramic Fast Reactor

To make the nuclear fuel cycle more economical and convenient, as well as prevent nuclear proliferation, the conceptual study of a simple high-temperature dry reprocessing of spent nuclear fuel (SNF) for a ceramic fast reactor is proposed in this paper. This simple high-temperature dry (HT-dry) reprocessing includes the Atomics International Reduction Oxidation (AIROX) process and purification method for rare-earth elements. After removing the part of fission products from SNF by a HT-dry reprocessing without fine separation, the remaining nuclides and some uranium are fabricated into fresh fuel which can be used back to the ceramic fast reactor. Based on the ceramic coolant fast reactor, we studied neutron physics of nuclear fuel cycle which consists operation of ceramic reactor, removing part of fission products from SNF and preparation of fresh fuels for many time. The parameters of the study include effective multiplication factor (Keff), beam density, and nuclide mass for different ways to remove the fission products from SNF. With the increase in burnup time, the trend of increasing 239Pu gradually slows down, and the trend of 235U gradually decreases and become balanced. For multiple removal of part of fission products in the nuclear fuel cycle, the higher the removal, the larger the initial Keff.

ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.

THE PRINCESS PROJECT: FROM DIFFERENTIAL TO INTEGRAL EXPERIMENTS

Following the shutdown of the CEA Valduc experimental facilities, where, for more than 50 years, IRSN used to perform experiments related to criticality safety, IRSN initiated a new project named PRINCESS (PRoject for IRSN Neutron physics and Criticality Experimental data Supporting Safety). The objective is to continue collecting experimental data necessary for the IRSN missions in nuclear safety. For this purpose, collaborations with various national and international laboratories have been established. The PRINCESS project covers various nuclear physics fields from nuclear data to criticality-safety and reactor physics providing information to both differential and integral data improvements.

TANGRA multidetector systems for investigation of neutron-nuclear reactions at the JINR Frank Laboratory of Neutron Physics

In the framework of TANGRA-project at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear research in Dubna (Russia), two experimental setups (Fig. 1) have been designed and tested for investigation of 14-MeV neutron-induced nuclear reactions on a number of important for nuclear science and engineering isotopes. As a source of 14-MeV “tagged” neutrons we are using the VNIIA ING-27 steady-state portable neutron generator with embedded in its vacuum tube 64-pixel charge-particle detector. The “Romashka” system is an array of up-to 24 hexagonal NaI(Tl)-crystal scintillation probes, while the “Romasha” array consists of 18 cylindrical BGO-crystal detectors of neutrons and gamma-rays. In addition to these detectors there is a HPGe gamma-ray spectrometer and a number of Stilbene detectors that can be added for high-resolution gamma-ray spectrometry and neutron-gamma detection. The main characteristics of the neutron-induced nuclear reaction products can be investigated by commissioning the detectors in suitable for these experiments’ geometries. Both setups can be used for doing basic and applied scientific research, because they permit simultaneously to measure the energy, angle and multiplicity distributions of gamma-rays and neutrons, produced in the competitive neutron-induced nuclear reactions (n, n’γ), (n,2n), (n, xnγ) and (n, f) in pure or complex substances.

Development and Verification of Multi-Physical Coupling Calculation Code for Typical PWR Core

Strong feedback phenomenon between the neutronics physical and thermal-hydraulic has an important impact on the design and safety analysis of pressured water reactor (PWR). In order to accurately simulate the strong coupling effect of neutron physics and thermal hydraulic in PWR, a reactor multi-physical coupling calculation code (ARMcc) is developed, in which the three-dimensional space-time neutron dynamic equation is solved by nodal expansion method (NEM) and nodal Green's function method (NGFM), the coolant temperature and fuel temperature are solved by single channel model and the cylinder heat conduction model respectively. The 3D IAEA benchmark, the 3D Langenbuch Maurer Werner (LMW) benchmark and NEACRP-L-335 benchmark are used to verify the neutronics model and coupling calculation solution ability respectively. The results show that: 1) the NEM and NGFM have high accuracy in solving the three-dimensional space-time neutron dynamics equation; 2) the results of neutronics and thermal-hydraulic coupling steady/transient calculation such as core relative power and fuel Doppler temperature are in good agreement with those of the NEACRP-L-335 benchmark, and the calculation accuracy is equivalent to similar software such as PARCS. Four coupled neutron physics and thermal hydraulic calculation modes are used to analyze the influence of different neutron physics calculation methods and thermal hydraulic calculation methods on the key parameters of PWR transient process in this paper. The results show that the mode of NGFM + FVM can more accurately simulate the core relative power peak and Doppler temperature.