Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors

In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

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Analyzing the Neutron Parameters in the Accelerator Driven Subcritical Reactor Using the Mixture of Molten Pb-Bi as Both Target and Coolant

Atoms ◽

10.3390/atoms9040095 ◽

2021 ◽

Vol 9 (4) ◽

pp. 95

Author(s):

Tien Tran Minh

Keyword(s):

Proton Beam ◽

Energy Levels ◽

Incident Proton ◽

The Core ◽

Neutron Multiplication ◽

Triga Mark Ii ◽

Effective Neutron ◽

Fuel Mixtures ◽

Subcritical Reactor ◽

Good Agreement

In this paper, the Accelerator Driven Subcritical Reactor (ADSR) was simulated based on the structure of the TRIGA-Mark II reactor by the MCNPX program. The proton beam interacts on the Pb-Bi molten target with various energy levels from 0.5 GeV to 2.0 GeV. The important neutron parameters to evaluate the operability of ADSR were calculated as: the neutron yields according to various thicknesses of the target and according to the energy of the incident proton beam; the effective neutron multiplication factor for various fuel mixtures, along with its stability for some fuel mixtures; the axial and radial distributions of the neutron flux along with the height and radius of the core. The obtained results had shown a good agreement in using Pb-Bi molten as the interaction target and coolant for ADSR.

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Analysis of Protected Accidental Transients in the EFIT Reactor With the RELAP5 Thermal-Hydraulic Code

Volume 3: Thermal Hydraulics; Instrumentation and Controls ◽

10.1115/icone16-48720 ◽

2008 ◽

Cited By ~ 1

Author(s):

Giacomino Bandini ◽

Maddalena Casamirra ◽

Francesco Castiglia ◽

Mariarosa Giardina ◽

Paride Meloni ◽

...

Keyword(s):

Proton Beam ◽

Thermal Power ◽

Heat Removal ◽

Reactor Power ◽

Lead Target ◽

Primary System ◽

Safety Behavior ◽

Forced Circulation ◽

Detailed Model ◽

Decay Heat

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of the European EUROTRANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside the reactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor in safe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system.

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Neutronic and Physical Characteristics of an Accelerator Driven System with a Lead-208 Coolant

Volume 3: Structural Integrity; Nuclear Engineering Advances; Next Generation Systems; Near Term Deployment and Promotion of Nuclear Energy ◽

10.1115/icone14-89153 ◽

2006 ◽

Author(s):

Georgy L. Khorasanov ◽

Anatoly P. Ivonov ◽

Anatoly I. Blokhin

Keyword(s):

Proton Beam ◽

Neutron Source ◽

Lead Isotope ◽

External Source ◽

Beam Current ◽

Thermal Capacity ◽

Multiplication Factor ◽

The Core ◽

Effective Multiplication Factor ◽

Effective Multiplication

In the paper a possibility of using a lead isotope, pure Pb-208, as a coolant for a subcritical core of 80 MW thermal capacity of the PDS-XADS type facility is considered. Calculations of neutronic characteristics were performed using Monte Carlo technique. The following initial data were chosen: an annular core with a target, as a neutron source, at its centre; the core coolant — Pb-208 (100%); a fuel — a mix of mono nitrides of depleted uranium and power plutonium with a small share of neptunium and americium; the target coolant — a modified lead and bismuth eutectic, Pb-208(80%)-Bi(20%); proton beam energy — 600 MeV; effective multiplication factor of the core under operation — Keff = 0.97; thermal capacity of the core — N = 80 MW. From calculations performed it follows that in using Pb-208 as the core coolant the necessary intensity of the external source of neutrons to deliver 80 MW thermal capacity is equal to S = 2.29−1017 n/s that corresponds to proton beam current Ip = 2.8 mA and beam capacity Pp = 1.68 MW. In using natural lead instead of Pb-208 as the core coolant, effective multiplication factor of the core in normal operating regime falls down to the value equal to Keff = 0.95. In these conditions multiplication of external neutrons in the core and thermal capacity of the subcritical core are below nominal by 1.55 times. For achievement the rated core power N = 80 MW it is required on ∼20–30% to increase the fuel loading and volume of the core, or by 1.55 times to increase intensity of the external source of neutrons. In the last case, the required parameters of the neutron source and of the corresponding proton beam are following: intensity of the neutron source S = 3.55·1017 n/s., beam current Ip = 4.32 mA, beam capacity Pp = 2.59 MW. To exploit the accelerator with the reduced proton beam current it will be required about 56 tons of Pb-208, as a minimum, for the core coolant. Charges for its obtaining can be recovered at the expense of the economy of the proton accelerator construction cost. In this case, the acceptable price of the lead isotope Pb-208 must be less than $2,860/kg.

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Characterization of the gamma flux in a tangential channel of the CENM TRIGA MARK II research reactor

EPJ Web of Conferences ◽

10.1051/epjconf/202022504030 ◽

2020 ◽

Vol 225 ◽

pp. 04030

Author(s):

A. Gruel ◽

D. Fourmentel ◽

C. El Younoussi ◽

B. El Bakkari ◽

Y. Boulaich ◽

...

Keyword(s):

Steady State ◽

Nuclear Data ◽

Neutron Field ◽

Photon Propagation ◽

The Core ◽

Thermal Maximum ◽

Gamma Field ◽

Triga Mark Ii ◽

Gamma Flux

The CNESTEN (National Center for Energy Sciences and Nuclear Technology, Morocco) operates a TRIGA Mark II reactor, which can reach a thermal maximum power at steady state of 2 MW. In reactors devoted to research and experiments, it is mandatory to characterize the neutron and photon fields in the irradiation positions. Together with a computational model of the core, it ensures the ability to reach the requested uncertainties when performing experiments, such as detectors testing, irradiation for hardening or nuclear data measurements. The neutron field of different irradiation positions has been characterized by dosimetry techniques and compared to the MCNP full model of the reactor. Preliminary photon propagation calculations are also performed with this model, but up to now, no experimental validation of the results exists. The aim of the newly set collaboration between CEA and CNESTEN is to characterize the gamma field of these positions. The first position investigated is the part of the NB1 tangential channel closest to the core. Among gamma measurements techniques, and according to the constraints arising from using this channel, it was chosen to use thermos- and optically stimulated luminescent detectors. This paper presents the experiments carried out in September 2018 as well as their results. Three detectors types were used: TLD400 (CaF2:Mn), TLD700 (7LiF:Mg,Ti) and OSLD (Al2O3:C). Measurements were performed in several steps: background measurements, transient measurements (divergence phase + SCRAM), and irradiation at steady state. In the end, these measurements will provide a dose as well as a gamma flux value for this position.

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