Institute Of Nuclear Energy Safety Technology Research Articles

Summary of the 2nd International Workshop on Gas-Dynamic Trap based Fusion Neutron Source (GDT-FNS)

The 2nd International Workshop on Gas-Dynamic Trap based Fusion Neutron Source (GDT-FNS) was held on 18-19 November 2019 in Novosibirsk, Russian Federation. It followed the First GDT-FNS Workshop, which took place in November 2018 in Hefei, China. The 2nd workshop was attended by thirty participants representing eight institutes and universities from five countries. Oral presentations were broadcast on the internet via the Zoom.us conferencing system and aroused the interest of scientists from Japan, China, USA, and Russian Federation. More than 20 presentations covered various stages of the ALIANCE project jointly proposed in 2019 by the Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences and Budker Institute of Nuclear Physics (BINP), Siberian Branch of Russian Academy of Sciences. The next GDT-FNS workshop is planned to be held in October 2020 in Hefei.

A full-digital multidimensional redundancy control system for high intensity D-T fusion neutron generator

High Intensity D–T fusion Neutron Generator (HINEG), developed by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS)•FDS Team, is an essential platform for advanced nuclear energy systems, which can be used for researches on nuclear technology and safety including the validation of neutronic method and software, validation and measurement of nuclear data, radiation shielding and protection, material irradiation damage and activation, neutronic performance of components, etc. In addition, the application of HINEG can be extended to nuclear medicine, radiotherapy, neutron imaging and other nuclear technology applications. A full-digital multidimensional redundancy control for HINEG system has been implemented by full-digital hardware of the Server/Client mode, and software infrastructure of data acquisition and process control designed by modular and multidimensional redundancy method. Either local users or remote users can execute all authorized control function by unified interface under the condition of radiation, ultra-high voltage and complex electromagnetic environment. With the application of HINEG Control System (HINEG CS), it is the first time for the neutron generator control system to realize full-digital and multidimensional redundancy control which can improve the reliability of neutron generator. HINEG CS has successfully been realized on HINEG, furthermore stability, reliability and good scalability have been well verified.

Measurement and Analysis of Neutron Leakage Spectra from Pb and LBE Cylinders with D-T Neutrons

For validating the current evaluated neutron data libraries, neutron leakage spectra from lead and lead bismuth eutectic (LBE) cylinders have been measured using an intense D-T pulsed neutron source with time-of-flight (TOF) method by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS). The measured leakage spectra have been compared with the calculated ones using Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) with the evaluated pointwise data of lead and bismuth processed from ENDF/B-VII.1, JEFF-3.1 and JENDL-4.0 libraries. This work shows that calculations of the three libraries are all generally consistent with the lead experimental result. For LBE experiment, the JEFF-3.1 and JENDL-4.0 calculations both agree well with the measurement. However, the result of ENDF/B-VII.1 fails to fit with the measured data, especially in the energy range of 5.5 and 7 MeV with difference more than 80%. Through sensitivity analysis with partial cross sections of 209 Bi in ENDF/B-VII.1 and JEFF, the difference between the measurement and the ENDF/B-VII.1 calculation in LBE experiment is found due to the neutron data of 209 Bi.

CLEAR‐S: an integrated non‐nuclear test facility for China lead‐based research reactor

Summary A project to develop the Accelerator Driven System (ADS) for nuclear waste transmutation has been launched by Chinese Academy of Sciences (CAS) since 2011 through three stages. China LEAd-based Reactor (CLEAR) was selected as the reference reactor for CASADS project, which means lead or lead-bismuth will be used as the primary coolant for reactor. Institute of Nuclear Energy Safety Technology (INEST), CAS, is one of the leading teams being involved in ADS project, and taking charge of the design and R&D activities of CLEAR series reactors. According to the research and development roadmap of CLEAR, a 10MWth lead-bismuth cooled pool-type research reactor named CLEAR-I coupled with a proton accelerator will be constructed at the first stage. In the second stage, China LEAd-based Engineering Demonstration Reactor (CLEAR-II) with 100MWth will be constructed. In the third stage, China LEAd-based Commercial Prototype Reactor (CLEAR-III) with 1000MWth will be constructed. In order to verify the key components and investigate the thermal-hydraulics phenomena for CLEAR-I and even for pool type lead-based reactor, an integrated multifunctional non-nuclear test facility named CLEAR-S will be built and commissioning before the end of 2016. Note that CLEAR-S will be used to test the 1:1 prototype components for CLEAR-I, such as primary pump, heat exchanger, control rod driven system and in-vessel refueling system, and to verify the design and safety analysis codes. In this paper, the design description and latest progress have been presented for CLEAR-S. © 2016 The Authors. International Journal of Energy Research Published by John Wiley & Sons Ltd.

Development strategy and conceptual design of China Lead-based Research Reactor

Chinese Academy of Sciences (CAS) launched an engineering project to develop an Accelerator Driven System (ADS) for nuclear waste transmutation since 2011, and China LEAd-based Reactor (CLEAR) proposed by Institute of Nuclear Energy Safety Technology (INEST) is selected as the ADS reference reactor. In this paper, the development strategy and conceptual design of China Lead-based Research Reactor are proposed. The Chinese ADS development program consists of three stages, and during the first stage, a 10MWth lead-based research reactor named CLEAR-I will be built with subcritical and critical dual-mode operation capability for validation of ADS transmutation system and lead cooled fast reactor technology. Major design principles of CLEAR-I are oriented at technology feasibility, safety reliability, experiment flexibility and technology continuity. Followed by the development strategy and design principles, CLEAR-I design options and conceptual design scenarios are presented.

Preliminary simulation of fuel dispersion in a Lead-Bismuth Eutectic (LBE)-cooled research reactor

In the severe accident analysis of liquid metal cooled reactors, it is important to understand the fuel dispersion behavior which may lead to re-criticality and influence on the heat removal under core damage conditions. For the safety investigation of a LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST), six cases with different parameters which influence the dispersion phenomenon in the coolant were simulated with the Neutronics and Thermal-hydraulics Coupled simulation program (NTC). Fuel dispersion transient results in six cases showed that the fuel porosity, coolant velocity pattern, fuel particle diameter and release position have a significant influence on fuel accumulation and redistribution. In these postulated cases, fuel sweep-out and the dispersion effects were dominating processes rather than fuel compaction and no re-criticality took place in the short term (1000 s). These results confirm that the reactor has certain resistance against severe accidents with re-criticality.

Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient.

Experiments and analysis on LBE steady natural circulation in a rectangular shape loop

For the purpose of investigating the thermal-hydraulic characteristics of lead-bismuth eutectic (LBE) natural circulation (NC), a LBE NC loop (NC-loop), experimental platform of the KYLIN-II facility was operated by the Institute of Nuclear Energy Safety Technology (INEST) in 2011. Experiments to characterize the thermal-hydraulic performance of HLM coolants undergoing NC were carried out at steady state to study the following issues: 1) the initiation of NC at different power levels; 2) the effects of input power and secondary loop operational characteristics on NC; 3) the capacity limits of NC and 4) a comparison of experimental data with theoretical predictions. The maximum NC capacity for LBE with a flow rate of 1.2 kg/s was obtained for a 24 kW heat pin corresponding to heating power input of 200 W/cm, which is higher than the linear power rating of the proposed China LEad Alloy Reactor-I (CLEAR-I). The experimental results at high heat pin input power were found to be in good agreement with the theoretical predictions.

Development status of CLAM steel for fusion application

The China low activation martensitic (CLAM) steel is being developed at the Institute of Nuclear Energy Safety Technology (INEST) under wide collaboration within China. Significant R&D work on CLAM steel was carried out to help make it suitable for industrial applications. The effect of refining processes and thermal aging on composition, microstructures and mechanical properties were investigated. Material properties before irradiation including impact, fracture toughness, thermal aging, creep and fatigue properties etc. were assessed. A series of irradiation tests in the fission reactor HFETR in Chengdu up to 2dpa and in the spallation neutron source SINQ in Paul Scherrer Institute up to 20dpa were performed. PbLi corrosion tests for more than 10,000h were done in the DRAGON-I and PICOLO loops. Fabrication techniques for a test blanket module (TBM) are being developed and a 1/3 scale TBM prototype is being fabricated with CLAM steel. Recent progresses on the development status of this steel are presented here. The code qualification of CLAM steel is under plan for its final application in ITER-TBM and DEMO in the future.