Feynman-α and Rossi-α formulas with spatial and modal effects

Feynman-α and Rossi-α formulas including multiple α-modes are derived for stochastic and continuous neutron sources. The presented formalism is further developed to achieve spatial correction factors for the single α-mode point kinetics representations of the Feynman-α and Rossi-α formulas. As a natural extension of the multiple α-mode formalism, delayed neutrons are included in the Feynman-α formula. The obtained formulas are validated experimentally in a strongly heterogeneous system obeying multiple α-modes, resulting in good agreement with the presented theoretical framework.     

Data decomposition method for full-core Monte Carlo transport–burnup calculation

Monte Carlo transport simulations of a full-core reactor with a high-fidelity structure have been made possible by modern-day computing capabilities. Performing transport–burnup calculations of a full-core model typically includes millions of burnup areas requiring hundreds of gigabytes of memory for burnup-related tallies. This paper presents the study of a parallel computing method for full-core Monte Carlo transport–burnup calculations and the development of a thread-level data decomposition method. The proposed method decomposes tally accumulators into different threads and improves the parallel communication pattern and memory access efficiency. A typical pressurized water reactor burnup assembly along with the benchmark for evaluation and validation of reactor simulations model was used to test the proposed method.The result indicates that the method effectively reduces memory consumption and maintains high parallel efficiency.     

Measurement and Monte Carlo simulation of γ-ray dose rate in high-exposure building materials

Natural radioactivity radionuclides in building materials, such as~(226)Ra,~(232)Th and~(40)K, cause indoor exposure due to their gamma-rays. In this research, in a standard dwelling room(5.0 m 9 4.0 m 9 2.8 m), with the floor covered by various granite stones, was set up to simulate the dose rates from the radionuclides using MCNP4 C code. Using samples of granite building products in Iran, activities of the~(226)Ra,~(232)Th and~(40)K were measured at 3.8–94.2, 6.5–172.2 and 556.9–1529.2 Bq kg~(-1),respectively. The simulated dose rates were26.31–184.36 n Gy h~(-1), while the measured dose rates were 27.70–204.17 n Gy h~(-1). With the results in good agreement, the simulation is suitable for any kind of dwelling places.     

Feynman-α Analysis for a Thermal Subcritical Reactor System Driven by an Unstable 14MeV Neutron Source

In a series of Feynman-α correlation measurements for a thermal Accelerator-Driven System (ADS) with 14MeV neutrons at the Kyoto University Critical Assembly (KUCA), an unstable accelerator condition such as a drift of beam current has been frequently observed. Neutron source instability caused by such unavoidable beam-current instability resulted in a divergent variance-to-mean ratio and, consequently, the correlation analysis failed. Nevertheless, we attempted to apply a difference-filtering technique to the correlation analysis to reduce the influence of the above instability. The present attempt resulted in consistent prompt-neutron decay constants with those obtained in a previous pulsed neutron experiment. The application of the filtering is expected to enhance the robustness of Feynman-α analysis against various instabilities of accelerator operation in actual ADS.     

An applied limit of the bunching method for the Feynman-α analysis

Feynman-α method is used as the representative method in reactor noise analysis for the criticality monitoring. Feynman-α analysis needs a large amount of measurement time in its original process, though many researchers use the bunching method and its derived methods for the experimental data processing to shorten the measurement time. However, the detailed characteristics and the application limit of the bunching method have not been researched and discussed enough. This paper shows a possibility that the Bunching method is a method to reduce the probability fluctuation with the Y value only in the appearance. Moreover, the criteria for determining that the Y value is not an accidental product are also provided in this paper.     

Science and Technology of Reactor Reactor Core Geometry Arrangement Research of Spent Fuel Critical Facility

With the development of nuclear power, the nuclear critical safety problem becomes more and more significant. The burnup credit technology has been applied to analysis of the nuclear critical safety. This has significantly enhanced the capacity of storing, transportation, reprocessing and improved the economy of back end fuel cycle. It is very important to carry out critical experiment on spent fuel for selecting calculation code packages and validating calculation methods using burnup credit technology. Before building the spent fuel critical experimental facility, massive detailed critical calculation should be performed.     

Primary study on holdup measurement of 235U in pipe using γ-ray spectrometry and Monte Carlo simulation

The pipe holdup measurement is very important for decommissioning nuclear facilities and nuclear-material control and accounting. The absolute detection efficiencies (εsp) of full-energy γ rays peak under different source density distribution function have been simulated using the Monte Carlo (MC) software, and the counting rates (n0) of the characteristic γ rays have been measured using the γ spectrometer followed by the calculation of the holdup. The holdup is affected by the energy of γ rays, distance at which they are detected, pipe material, thickness,and source distribution of pipe, especially source distribution at a short distance. The comparative test of 235U reference materials on the inner wall of Fe and A1 pipes (the total mass of 235U is 44.6 mg and 222.8 mg, respectively) have been accomplished using this method. The determined result of 235U is 43.2mg (U0.95rel=5.4%) and 216.2mg (U0.95rel= 3.2%), respectively, which are in accordance with the reference values.     

Primary study on holdup measurement of ^235U in pipe using γ-ray spectrometry and Monte Carlo simulation

The pipe holdup measurement is very important for decommissioning nuclear facilities and nuclear-material control and accounting. The absolute detection efficiencies （εsp） of full-energy γ rays peak under different source density distribution function have been simulated using the Monte Carlo （MC） software, and the counting rates （no） of the characteristic γ rays have been measured using the γ spectrometer followed by the calculation of the holdup. The holdup is affected by the energy of γ rays, distance at which they are detected, pipe material, thickness, and source distribution of pipe, especially source distribution at a short distance. The comparative test of ^235U reference materials on the inner wall of Fe and A1 pipes （the total mass of ^235U is 44.6 mg and 222.8 mg, respectively） have been accomplished using this method. The determined result of ^235U is 43.2mg （U0.95rel=5.4%） and 216.2mg （U0.95rel= 3.2%）, respectively, which are in accordance with the reference values.     

Comparison and physical interpretation of MCNP and TART neutron and γ Monte Carlo shielding calculations for a heavy-ion ICF system

Inertial confinement fusion (ICF) aims to induce implosions of D–T pellets to obtain a extremely dense and hot plasma with lasers or heavy-ion beams. For heavy-ion fusion (HIF), recent research has focused on “liquid-protected” designs that allow highly compact target chambers. In the design of a reactor such as HYLIFE-II [Fus. Techol. 25 (1984); HYLIFE-II Progress Report, UCID-21816, 4.82-100], the liquid used is a molten salt made of F¹⁰, Li⁶, Li⁷, Be⁹ (called flibe). Flibe allows the final-focus magnets to be closer to the target, which helps to reduce the focus spot size and in turn the size of the driver, with a large reduction of the cost of HIF electricity. Consequently the superconducting coils of the magnets closer to the D–T neutron source will potentially suffer higher damage though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced γ rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted.The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The technical nature of the design problem and the methodology followed were presented in a previous paper [Nucl. Instr. and Meth. A 464 (2001) 410] by summarizing briefly the results for the deposited energy distribution on the six focal magnets of a beam line. Now a comparison of the performances of the two codes TART98 [TART98: A Coupled Neutron-Photon 3-D Combinational Geometry Monte Carlo Transport Code, Lawrence Livermore National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997] and MCNP4B [MCNP – A General Monte Carlo N-Particle Transport Code, Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory] for two different configurations of the system is discussed, separating the n and γ contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary γ and of secondary γ generated by inelastically scattered or radiatively captured neutrons. The final conclusions indicate some guidelines and suggest possible improvements for the future neutronic shielding design for a HIF facility.     

Progress of Upgrading and Reconstruction Project for Phase Ⅰ and Phase Ⅱ of ESSALAM Reactor

正In the 1980s,an agreement was made by and between China and Algeria for the peaceful use o f nuclear energy and strengthening scientific and technical cooperation.China Institute of Atomic Energy(CIAE)provided a ssistance in the construction of Birine Nuclear Research Center(CRNB),Algeria.CRNB is a center for     

Investigation of high-temperature-resistant rhenium–boron neutron shields by experimental studies and Monte Carlo simulations

In this study,novel rhenium–boron neutronshielding high-temperature-resistant materials were designed.The considered samples,Re60–B40,Re58–B42,Re50–B50,and Re40–B60,with different concentrations of rhenium and boron were investigated to elucidate their neutron-shielding performances,and compare them with well-known neutron-shielding materials such as the 316 LN quality nuclear steel.In addition to the experimental studies,Monte Carlo simulations were performed using the FLUKA and GEANT4 codes,where 4.5-MeV neutrons emitted by a ~(241)Am–Be source were employed.Experimental equivalent dose rates,simulated track lengths,energy balances,and neutron mass absorption cross sections were discussed in detail.     

Comparative analyses of coated and composite UN fuel – Monte Carlo based full core LWR study

Longer continuous operation of a nuclear reactor leads to higher availability of nuclear power plant, entailing economic gain. In this context the option being explored, recently, is the use of higher density fuels as compared to current fuels i.e. UO₂. Uranium mono nitride (UN) fuel is one of the options being explored in this regard. UN fuel has been used in nuclear industry for a long time with fast reactor option. More recently, studies have targeted its use in Light Water Reactor (LWR) environment. The main problem with using UN fuel in LWR is its potential reaction with water which produces hydrogen. Researchers have proposed different approaches to overcome this problem.One option is the use of coatings around UN fuel pellet to avoid the direct contact of water with fuel. The second option is use of a secondary phase (10 vol. percent) like ZrO₂ which can make an oxide layer in case of contact with water and protect the main phase, Uranium mono nitride (80 vol. percent). Remaining 10 vol. percent is considered to be consumed by porosity. This study aims at comparison of neutron physics behavior of both these options in LWR environment. The upshot of the study is to quantify the impact of adding layers or secondary phase with respect to pure/complete UN fuel. To study these effects, two theoretical densities i.e. 95% and 80% for UN fuel are chosen for analyses. To avoid the problem of C-14 production from N-14, fuels studied are considered to be having 100% N-15.A validated model of Benchmark for Evaluation and Validation of Reactor Studies (BEAVRS) is used to perform all the analyses. Integral neutron physics parameters like neutron energy spectra, Radial Peaking Factors, Axial Peaking Factor, Doppler coefficient of reactivity, Isothermal Coefficient of reactivity and Temperature defect, Control Rod worth and excess reactivity for whole core are compared at Beginning of Cycle (BoC). Burnup obtained by different fuel option is also compared. Consistent with pin-cell based earlier findings, this full 3D analyses with UN based fuel shows noticeable spectral hardening leading to decrease in the value of control rod worth and less negative Doppler coefficient of reactivity while power peaking factors remain mostly unchanged. The economic advantage of switching to UN based fuels is expected when UN fuel above 80% TD is used. Approximately 19% increase in Equivalent Full Power Days (EFPDs) is witnessed by using 95% TD UN based fuels.     

Benchmarking of the 3-D CAD-based Discrete Ordinates code “ATTILA” for dose rate calculations against experiments and Monte Carlo calculations

Shutdown dose rate (SDDR) inside and around the diagnostics ports of ITER is performed at PPPL/UCLA using the 3-D, FEM, Discrete Ordinates code, ATTILA, along with its updated FORNAX transmutation/decay gamma library. Other ITER partners assess SDDR using codes based on the Monte Carlo (MC) approach (e.g. MCNP code) for transport calculation and the radioactivity inventory code FISPACT or other equivalent decay data libraries for dose rate assessment. To reveal the range of discrepancies in the results obtained by various analysts, an extensive experimental and calculation benchmarking effort has been undertaken to validate the capability of ATTILA for dose rate assessment. On the experimental validation front, the comparison was performed using the measured data from two SDDR experiments performed at the FNG facility, Italy. Comparison was made to the experimental data and to MC results obtained by other analysts. On the calculation validation front, the ATTILA's predictions were compared to other results at key locations inside a calculation benchmark whose configuration duplicates an upper diagnostics port plug (UPP) in ITER. Both serial and parallel version of ATTILA-7.1.0 are used in the PPPL/UCLA analysis performed with FENDL-2.1/FORNAX databases. In the FNG 1st experimental, it was shown that ATTILA's dose rates are largely over estimated (by ～30–60%) with the ANSI/ANS-6.1.1 flux-to-dose factors whereas the ICRP-74 factors give better agreement (10–20%) with the experimental data and with the MC results at all cooling times. In the 2nd experiment, there is an under estimation in SDDR calculated by both MCNP and ATTILA based on ANSI/ANS-6.1.1 for cooling times up to ～4 days after irradiation. Thereafter, an over estimation is observed (～5–10% with MCNP and ～10–15% with ATTILA). As for the calculation benchmark, the agreement is much better based on ICRP-74 1996 data. The divergence among all dose rate results at ～11 days cooling time is no more than 15% among all participants.     

Dynamic scenario quantification for level 2 PRA of sodium-cooled fast reactor based on continuous Markov chain and Monte Carlo method coupled with meta-model of thermal–hydraulic analysis

In probabilistic risk assessment (PRA), an event tree (ET) methodology is widely used to quantify accident scenarios which result in core damage and fission products release. However, the current approach using the ET methodology is not applicable to evaluate dynamic characteristics of accident progression, when the accident progression is time-dependent and headings in the ET have inter-dependency between events. Thus, a dynamic approach of accident scenario quantification is necessary to evaluate more realistic PRA.

This research addressed this need by developing a dynamic scenario quantification method for the level 2 PRA by coupling of Continuous Markov chain and Monte Carlo (CMMC) method and a plant thermal–hydraulic analysis code for a sodium-cooled fast reactor (SFR).

The CMMC method is applied to protected loss of heat sink (PLOHS) accident of the SFR to analyze dynamic scenario quantifications. The coupling method requires heavy computational cost and it makes difficult to quantify the whole accident scenarios by comparing the results from existing plant state analysis codes. Thus, a meta-analysis coupling method is proposed to obtain dynamic scenario quantifications with reasonable computational cost. Also, a categorizing method is used to depict analytical results in a transparent manner.     

CTOF measurements and Monte Carlo analyses of neutron spectra for the backward direction from an iron target irradiated with 400–1200 MeV protons

A calorimetric-time-of-flight (CTOF) technique was used for real-time, high-precision measurement of the neutron spectrum at an angle of 175° from the initial proton beam direction, which hits a face plane of a cylindrical iron target of 20 cm in diameter and 25 cm thick. A comparison was performed between the neutron spectra predicted by the MARS and the MCNPX codes and that measured for 400, 600, 800, 1000 and 1200 MeV protons.