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[en] Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this work represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.
[en] Highlights: • A containment design is proposed for the I2S-LWR, including passive safety systems. • Assessment of the safety systems demonstrates indefinite passive cooling. • The impact of the ADS valves and accumulator on plant response is investigated. - Abstract: The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel reactor design concept which aims at delivering an electric power output level comparable to that of large LWRs (approximately 1000 MWe), while at the same time achieving an overall level of safety that is enhanced with respect to large Generation III+ LWRs. One of the main safety goals is to achieve indefinite cooling following design basis accidents (DBAs) using atmosphere air as ultimate heat sink. In order to accommodate these goals, the I2S-LWR incorporates several innovative safety features, including an integral Reactor Pressure Vessel (RPV), enhanced passive Decay Heat Removal (DHR) systems and several containment passive cooling systems. The present work is focused on a passive and reliable containment design, which plays a significant role in LOCA scenario and is the last boundary to prevent the release of radioactivity to the environment. In this paper, several innovative passive systems located in the I2S-LWR containment are proposed, including Core Make-up Tank (CMT), Accumulator (ACC), Passive Suppression System (PSS), Automatic Depressurization System (ADS), Passive Containment Cooling System (PCCS) and Passive Reactor Cavity Cooling System (PRCCS). The best-estimate thermal-hydraulic code RELAP5 has been used to model the I2S-LWR RPV and containment passive systems. The inadvertent opening of PORV (Power Operated Relief Valve) accident scenario has been simulated in order to study the containment response, including the coupling between RPV and containment. The results show that, through the activation of the ADS, pressure equilibrium between containment and reactor pressure vessel is achieved, while maintaining the reactor core covered at all times. The containment passive cooling systems ensure that the containment pressure remains at acceptable levels throughout the transient. A sensitivity study on the number of operating ADS valves, and on the availability of accumulators is also performed.
[en] Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of "2"4"2Pu, "2"3"2Th, natural uranium, and 93% enriched "2"3"5U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.
[en] In transport theory, adjoint-based partial derivatives with respect to mass density are constant-volume derivatives. Likewise, adjoint-based partial derivatives with respect to surface locations (i.e., internal interface locations and the outer system boundary) are constant-density derivatives. This study derives the constant-mass partial derivative of a response with respect to an internal interface location or the outer system boundary and the constant-mass partial derivative of a response with respect to the mass density of a region. Numerical results are given for a multiregion two-dimensional (r-z) cylinder for three very different responses: the uncollided gamma-ray flux at an external detector point, keff of the system, and the total neutron leakage. Finally, results from the derived formulas compare extremely well with direct perturbation calculations.
[en] A hidden Markov model method proposed earlier for passive acoustic leak detection in sodium fast reactor systems has been improved in order to clarify how to set all free model parameters and to allow smaller amounts of training data. The method is based on training the model on known background noise only and optimizing its free model parameters by a parametric study of detection performance for synthetic noises superposed onto the same background. This means that the method is not assuming any knowledge on the noise to be detected and may be used as a general fault detection method, even if the application envisaged here is leak detection for sodium fast reactors. Using recordings of background noise as well as from argon injection tests performed at full power in the Phenix sodium fast reactor plant, it is estimated that the resulting method will detect leak-like deviations from the background noise with a detection delay of a few seconds, a false alarm rate close to 108 per second and at signal-to-noise ratio conditions at least corresponding to an additive signal at 10 dB. The method is one-channel, i.e. using input from one single acoustic sensor only. (authors)
[en] A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects, stemming from lattice atoms thermal motion and accounts for it within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler Reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to -10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes the results done on this topic to date
[en] Permanent disposal of low- and intermediate-level radioactive wastes in the subterranean environment has been the preferred method of many countries, including Korea. A safety issue after the closure of a geological repository is that biodegradation of organic materials due to microbial activities generates gases that lead to overpressure of the waste containers in the repository and its disintegration with the release of radionuclides. As part of an ongoing large-scale in situ experiment using organic wastes and groundwater to simulate geological radioactive waste repository conditions, we investigated the geochemical alteration and microbial activities at an early stage (~63 days) intended to be representative of the initial period after repository closure. The increased numbers of both aerobes and facultative anaerobes in waste effluents indicate that oxygen content could be the most significant parameter to control biogeochemical conditions at very early periods of reaction (<35 days). Accordingly, the values of dissolved oxygen and redox potential were decreased. The activation of anaerobes after 35 days was supported by the increased concentration to ~50 mg L-1 of ethanol. These results suggest that the biogeochemical conditions were rapidly altered to more reducing and anaerobic conditions within the initial 2 months after repository closure. Although no gases were detected during the study, activated anaerobic microbes will play more important role in gas generation over the long term
[en] The conclusions reached in the recent paper by G.Th. Analytis (Ann. Nucl. Energy, vol. 8, p. 349, 1981) are not considered to be new. Several previous works on neutron noise propagation and BWR neutron noise in heterogeneous media are cited. Taken together they emphasise the importance of heterogeneous effects in understanding the processes by which noise sources propagate in the reactor. (author)
[en] Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger and the Natural Draft Heat Exchanger. In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, and to keep the DRACS ready for activation, if needed, during accidents. To help with the design and thermal performance evaluation of the DRACS, a computer code using MATLAB has been developed. This code is based on a one-dimensional formulation and its principle is to solve the energy balance and integral momentum equations. By discretizing the DRACS system in the axial direction, a bulk mean temperature is assumed for each mesh cell. The temperatures of all the cells, as well as the mass flow rates in the DRACS loops, are predicted by solving the governing equations that are obtained by integrating the energy conservation equation over each cell and integrating the momentum conservation equation over each of the DRACS loops. In addition, an intermediate heat transfer loop equipped with a pump has also been modeled in the code. This enables the study of flow reversal phenomenon in the DRACS primary loop, associated with the pump trip process. Experimental data from a High-Temperature DRACS Test Facility (HTDF) are not available yet to benchmark the code. A preliminary code validation is performed by using natural circulation experimental data available in the literature, which are as closely relevant as possible. The code is subsequently applied to the HTDF that is under construction at the Ohio State University.
[en] In the international neutron libraries, the behavior with the energy of the neutron cross sections of hydrogen in light water depends on the Thermal Scattering Laws tabulated in terms of S(α, β). For the Joint Evaluated Fission and Fusion library (JEFF), Mattes and Keinert have established Thermal Scattering Laws by using the LEAPR module of the NJOY code. However, uncertainties on the corresponding S(α, β) were never reported. Such missing information was recently calculated with the nuclear data code CONRAD by determining the covariances between the model parameters involved in LEAPR. The obtained uncertainties were propagated to reactivity coefficients calculated for critical assemblies operating in 'cold' conditions (temperature below 80 C) and for PWR in 'hot' operating conditions (300 C). For the integral benchmarks investigated in this work, we found that the uncertainty on the calculated keff, due to the S(α, β) uncertainties, is close to ±130 pcm at room temperature and ±50 pcm at 300 C. (authors)