Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

A perturbation method is proposed to obtain the effective delayed neutron fraction βeff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Ap of the sample in βeff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation βeff=dk/△ρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average βeff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for fleer can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 23SU froin G.R. Keepin＇s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2.

Application of Monte Carlo method to calculate the effective delayed neutron fraction in molten salt reactor

Delayed neutron loss is an important parameter in the safety analysis of molten salt reactors. In this study,to obtain the effective delayed neutron fraction under flow condition, a delayed neutron precursor transport was implemented in the Monte Carlo code MCNP. The moltensalt reactor experiment(MSRE) model was used to analyze the reliability of this method. The obtained flow losses of reactivity for 235 U and 233 U fuels in the MSRE are223 pcm and 100.8 pcm, respectively, which are in good agreement with the experimental values(212 pcm and100.5 pcm, respectively). Then, six groups of effective delayed neutron fractions in a small molten salt reactor were calculated under different mass flow rates. The flow loss of reactivity at full power operation is approximately105.6 pcm, which is significantly lower than that of the MSRE due to the longer residence time inside the active core. The sensitivity of the reactivity loss to other factors,such as the residence time inside or outside the core and flow distribution, was evaluated as well. As a conclusion,the sensitivity of the reactivity loss to the residence time inside the core is greater than to other parameters.

Neutronic analysis of silicon carbide cladding accident-tolerant fuel assemblies in pressurized water reactors

In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.

Study on neutronics design of ordered-pebble-bed fluoride-salt- cooled high-temperature experimental reactor

This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed,which can keep core stability and meet the space requirements for thermal hydraulics and neutronics measurements.Overall, objectives of the core include inherent safety and sufficient excess reactivity providing 120 effective full power days for experiments. Considering the requirements above, the reactive control system is designed to consist of 16 control rods distributed in the graphite reflector. Combining the large control rods worth about 18000–20000 pcm, molten salt drain supplementary means(-6980 to -3651 pcm) and negative temperature coefficient(-6.32 to -3.80 pcm/K) feedback of the whole core, the reactor can realize sufficient shutdown margin and safety under steady state. Besides, some main physical properties, such as reactivity control, neutron spectrum and flux, power density distribution, and reactivity coefficient,have been calculated and analyzed in this study. In addition, some special problems in molten salt coolant are also considered, including ~6Li depletion and tritium production.

Neutronic Analysis of Generic Heavy Water Research Reactor Core Parameters to Use Standard Hydride Fuel

This research presents neutronic calculations for heavy water research reactor core substituting hydride fuel for uranium dioxide fuel. The aim of this research is feasibility analysis of reactor utilization with its original design using a new proposed fuel and changing the coolant and moderator circuit to light water. The required group constants for the CITATION code will be calculated using WIMSD-4 code. Neutronic calculations such as multiplication factors, radial and axial power peaking factor and fuel burn-up calculations are carried out by the CITATION code.

Real-time wide-range neutron flux monitor for thorium-based molten salt reactor

A novel full-digital real-time neutron flux monitor(NFM) has been developed for thorium-based molten salt reactor(TMSR).The system is based on the highspeed,parallel,and pipeline processing of the field programmable gate array as well as the high-stability controller area network platform.A measurement range of 10~8 counts per second is achieved with a single fission chamber by utilizing the normalization of the count and Campbell algorithms.With the advantages of using the measurement range,system integrity,and real-time performance,digital NFM has been tested in the Xi'an pulsed reactor fission experiments and was found to exhibit superior experimental performance.

Radiation Effect of Neutrons in a Reactor on Polyurethane

The radiation effect of neutrons in a reactor on polyurethane was studied.The gases produced by irradiated samples were analyzed by gas chromatography,and the dynamic mechanical and compression properties of the samples were also studied.The positron annihilation lifetime of irradiated samples was measured at room temperature in vacuum.The experimental results indicate that gas chromatography is a powerful tool to quantitatively analyze the gas products from neutron-irradiated polyurethane and characterizes the chemical changes in the sample.And the changes in microstructure determined from the PAL correlate well with the measurements of the mechanical properties by dynamic mechanical analysis（DMA）.

Xenon poisoning calculation code for miniature neutronsource reactor (MNSR)

In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR’s xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.

Unfolding the measured neutron spectra in the irradiation chamber of the UZrH reactor using iterative method

In the procedure of neutron fluence measurement in the whole energy range (10-4 eV^18 MeV), in the irra- diation chamber of a UZrH reactor, the neutron energy spectra are unfolded using the method of minimizing directed divergence and SAND-II, which are used broadly at home and abroad. These methods belong to the iterative methods. In this article, the procedure of the spectra unfolding using the two methods is described in detail. The neutron spec- trum distribution unfolded by the two methods agree well with each other. In the end, the major differences of the two iterative methods are compared with each other, and the main factors affecting the accuracy of the spectra unfolding with the iterative method are discussed.

Influence of the Neutron Flux Characteristic Parameters in the Irradiation Channels of Reactor on NAA Results Using k₀-Standardization Method

An approximation method using to estimate the influence of the uncertainties of the neutron flux characteristic parameters in the irradiation positions on the NAA results using k0-standardization technique was presented. Those are the epithermal reactor neutron spectrum shape-factor α, the effective resonace energy Ε for a given nuclide and the thermal to epithermal neutron flux ratio f. The method is applied to estimate the effect of the uncertainties in the determination of α Ε, and f on final NAA results for some irradiation channels of the Dalat reactor. It also shows that presented method is suitable in practical use for the estimation of the errors due to the uncertainty of the neutron flux characteristic parameters at the irradiation position.

Development of 24 and 59 keV Filtered Neutron Beams for Neutron Capture Experiments at Dalat Research Reactor

External filtered neutron beams have been developed at the horizontal radial channels No. 4 of Dalat research reactor. In the material composition of the neutron filters, the primary material components of Iron, Aluminum, Nickel and Vanadium are used to obtain the mono-energetic neutron beams of 24 and 59 keV, with low level of Gamma and slow neutron background. A computer code and Monte-Carlo simulation technique were applied to optimize the filter configurations and to deduce the neutron energy distributions in the filtered beams. A hydrogen-filled proton recoil detector and the activation method with Gold foils were used to measure the neutron energy spectrum and flux of each beam at sample position. The results of experimental neutron fluxes are 6.1 × 105 and 5.3 × 105 n/cm2/s for 24 and 59 keV beams, respectively.

Enhancement of neutron irradiation uniformity for the CFBR-Ⅱ fast burst reactor with a biaxial rotational technique

A biaxial rotational technique is proposed to improve the neutron irradiation uniformity for a large sample,and the theoretical method is established to predict and design the main parameters. The technique used a device to rotate the target sample around two perpendicular axes simultaneously. Numerical calculations found that the lowest common multiple of the two angular speeds should be large enough to improve the uniformity,and the minimal experimental time should be no less than 600 s. For a three-dimensional sample with a size of 20 cm × 12 cm × 14 cm, the maximal non-uniform neutron irradiation factor of the sample is mainly determined by the distance between the center of the sample and of the point neutron source. It was computed to be less than 10% when the distance was no less than 34 cm. Experiments were carried out on the CFBR-II reactor and the experimental results were in good accordance with the theoretical analysis. As a result, the theoretical conclusions given above are reasonable and of reference value for the design of future irradiation experiments.

Analysis of displacement damage effects on bipolar transistors irradiated by spallation neutrons

Displacement damage induced by neutron irradiation in China Spallation Neutron Source(CSNS) is studied on bipolar transistors with lateral PNP, substrate PNP, and vertical NPN configurations, respectively. Comparison of the effects on different type transistors is conducted based on displacement damage factor, and the differences are analyzed through minority carrier lifetime calculation and structure analysis. The influence of CSNS neutrons irradiation on the lateral PNP transistors is analyzed by the gate-controlled method, including the oxide charge accumulation, surface recombine velocity,and minority carrier lifetime. The results indicate that the total ionizing dose in CSNS neutron radiation environment is negligible in this study. The displacement damage factors based on 1-MeV equivalent neutron flux of different transistors are consistent between Xi’an pulse reactor(XAPR) and CSNS.

Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors

Accident-tolerant fuel(ATF)has attracted considerable research attention since the 2011 Fukushima nuclear disaster.To improve the accident tolerance of the fuel-cladding systems in the current light-water reactors,it is proposed to develop and deploy(1)an enhanced Zrbased alloy or coated zircaloy for the fuel cladding,(2)alternative cladding materials with better accident tolerance,and(3)alternative fuels with enhanced accident tolerance and/or a higher U density.This review presents the features of the current UO2-zircaloy system.Different techniques and characters to develop coating materials and enhanced Zr-based alloys are summarized.The features of several selected alternative fuels and cladding materials are reviewed and discussed.The neutronic evaluations of alternative fuel-cladding systems are analyzed.It is expected that one or more types of ATF-cladding systems discussed in the present review will be implemented in commercial reactors.

Feasibility analysis of 60Co production in pressurized water reactors

The radioactive isotope 60Co is used in many applications and is typically produced in heavy water reactors.As most of the commercial reactors in operation are pressurized light water reactors(PWRs),the world supply of high level radioactive cobalt would be greatly increased if 60Co could be produced in them.Currently,60Co production in PWRs has not been extensively studied;for the 59Co(n,c)60Co reaction,the positioning of 59Co rods in the reactor determines the rate of production.This article primarily uses the models of 60Co production in Canadian CANDU power reactors and American boiling water reactors;based on relevant data from the pressurized water Daya Bay nuclear power plant,a PWR core model is constructed with the Monte Carlo N-Particle Transport Code;this model suggests changes to existing fuel assemblies to enhance 60Co production.In addition,the plug rods are replaced with 59Co rods in the improved fuel assemblies in the simulation model to calculate critical parameters including the effective multiplication factor,neutron flux density,and distribution of energy deposition.By considering different numbers of 59Co rods,the simulation indicates that different layout schemes have different impact levels,but the impact is not large.As a whole,the components with four 59Co rods have a small impact,and the parameters of the reactor remain almost unchanged when four 59Co rods replace the secondary neutron source.Therefore,in theory,the use of a PWR to produce 60Co is feasible.

Instrumental neutron activation analysis of beachrock samples of South East Coast of Tamilnadu,India

Element profiles of some beach rock samples collected from South East Coast of Tamilnadu, India have been determined using single comparator method of INAA. The geo-chemical behavior of the elements in the region is discussed. The irradiations were done at thermal neutron flux of ~ 1011 cm-2·s-1 at 20kW power using Kalpakkam Mini Reactor (KAMINI), IGCAR, Kalpakkam, Tamilnadu, India. Around 19 elements have been determined from 15 samples by high-resolution gamma spectrometry. The accuracy and precision were evaluated by assaying the irradiated Standard Reference Material (SRM 1646a Estuarine sediment) and were found to be in good agreement with certified values.

Development of a three dimension multi-physics code for molten salt fast reactor

Molten Salt Reactor(MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum(GIF).The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors.In the present paper,a new coupling model is presented that physically describes the inherent relations between the neutron flux,the delayed neutron precursor,the heat transfer and the turbulent flow.Based on the model,integrating nuclear data processing,CAD modeling,structured and unstructured mesh technology,data analysis and visualization application,a three dimension steady state simulation code system(MSR3DS) for the can-type molten salt fast reactor is developed and validated.In order to demonstrate the ability of the code,the three dimension distributions of the velocity,the neutron flux,the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter(MOSART) using this code.The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor.Furthermore,the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion.

Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008.In this paper,we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the robe were obtained using the MCNP code without influence of y-ray and electronic-noise.The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated.The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI.The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques.And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.

Dynamic linear calibration method for a wide range neutron flux monitor system in ITER

As a key part of the diagnosis system in the International Thermonuclear Experimental Reactor(ITER),the neutron flux monitor(NFM),which measures the neutron intensity of the fusion reaction,is a Counting-Campbelling system with a large dynamic counting range.A dynamic linear calibration method is proposed in this paper to solve the problem of cross-over between the different counting and Campbelling channels,and improve the accuracy of the cross-calibration for long-term operation.The experimental results show that the NFM system with the dynamic linear calibration system can obtain the neutron flux of the fusion reactor in real time and realize the seamless measurement area connection between the two channels.