Neutronic performance of the MEGAPIE spallation target under high power proton beam

The MEGAPIE project, aiming at the construction and operation of a megawatt liquid lead-bismuth spallation target, constitutes the first step in demonstrating the feasibility of liquid heavy metal target technologies as spallation neutron sources. In particular, MEGAPIE is meant to assess the coupling of a high power proton beam with a window-concept heavy liquid metal target. The experiment has been set at the Paul Scherrer Institute (PSI) in Switzerland and, after a 4-month long irradiation, has provided unique data for a better understanding of the behavior of such a target under realistic irradiation conditions. A complex neutron detector has been developed to provide an on-line measurement of the neutron fluency inside the target and close to the proton beam. The detector is based on micrometric fission chambers and activation foils. These two complementary detection techniques have provided a characterization of the neutron flux inside the target for different positions along its axis. Measurements and simulation results presented in this paper aim to provide important recommendations for future accelerator driven systems (ADS) and neutron source developments.     

Neutronic analysis of deuteron-driven spallation target

Deuteron-driven spallation targets have garnered attention recently because they can provide high-energy neutrons to transmute long-lifetime fission products. I...     

Coupled fluid/structure analyses of the MEGAPIE spallation source target during transients

The paper describes some R&D activities undertaken in support of the design and safe operation of MEGAPIE (MEGAwatt PIlot Experiment) spallation source target, which is scheduled to be irradiated by a proton beam in the SINQ facility at the Paul Scherrer Institute in 2006. The target material is lead bismuth eutectic (LBE), which also acts as the primary coolant. As a consequence of the spallation reactions, about 600 kW of heat would be deposited in the target during operation, and considerable R&D effort is being expended to demonstrate continuing coolability and structural integrity under a variety of operational and abnormal conditions. The paper gives three examples of transient analyses carried out as part of the safety assessment of the target: (1) a beamline trip and recovery; (2) failure of the primary electro-magnetic pump (EMP); (3) failure of the secondary EMP (used to cool the base of the target). The study involves the simultaneous application of a system-analysis code, in our case a version of RELAP5, a computational fluid dynamics (CFD) tool (CFX-4), and a structural analysis code (ABAQUS). The RELAP5 code is used to provide transient boundary conditions for a localized conjugate heat transfer analysis of the lower target region, undertaken using CFD, and includes the feed-back effects arising from the secondary cooling and control systems. A conjugate heat transfer problem is then solved using CFD, which provides time-dependent thermal and flow data within the LBE, together with the thermal and mechanical loads to the target structures. Finally, an in-house interface program is employed to transfer mesh geometry, model topology and (time-dependent) thermal/mechanical data to enable stress analysis of the principal lower-target structural components to be performed. It is demonstrated that none of the transients considered result in critical stress conditions occurring in the target components, but that further operation is not recommended unless both pumps are fully operational.     

Assessment of the power deposition on the MEGAPIE spallation target using the GEANT4 toolkit

This work aims at evaluating the reliability of the GEANT4(GEometry ANd Tracking 4) Monte Carlo(MC) toolkit in calculating the power deposition on the Megawatt Pilot Experiment(MEGAPIE), the first liquid–metal spallation target worldwide. A new choice of codes to study and optimize this target is provided. The evaluation of the GEANT4 toolkit is carried out in comparison with the MCNPX and FLUKA MC codes. The MEGAPIE is an international project led by the Paul Scherrer Institute in Switzerland. It aims to demonstrate the safe operation of an intense neutron source to power the next generation of nuclear reactors, accelerator-driven systems(ADSs). In this study, we used the GEANT4 MC toolkit to calculate the power deposited by fast protons on the MEGAPIE target.The calculation focuses on several structures and regions.The predictions of our calculations were compared and discussed with that of the MCNPX and FLUKA codes,adopted by the MEGAPIE project. The comparison shows that there is a very good agreement between our results and those of the reference codes.     

Neutronic limits in infinite target mediums driven by high energetic protons

This study presents the neutronic behavior of integral data in an infinite target medium driven by an isotropic point source of 1000 MeV incident proton. Lead–bismuth eutectic, mercury, tungsten, uranium, thorium, chromium, copper and beryllium are considered as the target material because of their favorable spallation-neutron production characteristics. Furthermore, the calculations are performed for also dual mixture of some of them. In order to be able to simulate the infinite target medium by eliminating the spatial dependence, a spherical target is considered, and its radius is increased gradually up to adequate radius ensuring the infinite target medium. In this way, the radius value ensuring the maximum neutron leakage out of the target would be determined. Numerical calculations were performed with the high-energy Monte Carlo code MCNPX in coupled neutron and proton mode using the LA150 library. The mixing of the LBE with a solid target material (such as W, U and Th) lowers significantly the target radius ensuring the maximum neutron leakage.     

Experience from the post-test analysis of MEGAPIE

The (MEGAWatt Pilot Experiment) MEGAPIE target was successfully irradiated in 2006 at the SINQ facility of the Paul Scherrer Institut. During the irradiation a series of measurements to monitor the operation of the target, the thermal hydraulics behavior and the neutronic and nuclear aspects, has been performed. In the post-test analysis phase of the project, the data were analyzed and important information relevant to accelerator-driven systems (ADS) was gained, in particular: (i) from the operation of the target several recommendations concern the simplification of the system and the improved reliability; (ii) data from the thermal hydraulic measurements have offered the opportunity to validate the codes used in the design phase; (iii) the neutronic analysis confirm the high performance of a liquid metal target and the importance of the delayed neutron measurements in an ADS target; (iv) the nuclear measurements of the gas released gave the opportunity to validate the codes used during the design phase and provided indications for the operation. From the results in these different domains recommendations to further development of ADS and heavy liquid metal targets are discussed.     

Neutronic analysis for the IFMIF target and test cell using a new CAD-based geometry model

This work presents neutronic analyses to support the IFMIF target and test cell (TTC) design in the framework of the Broader Approach activities. A very detailed Monte Carlo geometry model of IFMIF TTC based on the modular TTC concept was prepared directly from a CAD model by using the McCad conversion software which has been developed at KIT. The Monte Carlo code McDeLicious, which is an enhancement to MCNP5, was utilized and nuclear heating, displacement damage and gas production rates in the TTC vessel wall were calculated. The calculation result shows that there are two prominent peaks; downstream of the test modules due to the high energy neutron contribution for gas productions and upstream due to neutron back-streaming along the beam ducts. The result suggests it is very important in the neutronic analysis to consider the detailed configuration of TTC and test modules. The dose rate distribution during operation has been assessed for the rooms adjacent to TTC across thick surrounding walls. The necessary thickness for the shielding walls has been examined. The result demonstrates the substantial improvement in the shielding capability for the top access cell with the present TTC design.     

MEGAPIE at SINQ - The first liquid metal target driven by a megawatt class proton beam

The lead-bismuth liquid metal target MEGAPIE (MEGAwatt Pilot Experiment) was operated at the Swiss Spallation Neutron Source SINQ starting mid-August 2006, for a scheduled irradiation period until 21st of December 2006. The continuous (51 MHz) 590 MeV proton beam hitting the target reaches routinely an average current of ∼1300 μA, corresponding to a beam power 0.77 MW. This article illustrates the main features of the target and the ancillary systems specially needed for the liquid metal target operation. Further, the operational experiences made with this target during start-up and routine operation are summarized, besides the general performance highlighting new beam and target safety devices, and last but not least the neutronic efficiency in relation to the previously operated solid lead target.     

Turbulent heat mixing of a heavy liquid metal flow in the MEGAPIE target geometry—The heated jet experiment

The MEGAPIE target installed at the Paul–Scherrer Institute is an example of a spallation target using eutectic liquid lead–bismuth (Pb⁴⁵Bi⁵⁵) both as coolant and neutron source. An adequate cooling of the target requires a conditioning of the flow, which is realized by a main flow transported in an annular gap downwards, u-turned at a hemispherical shell into a cylindrical riser tube. In order to avoid a stagnation point close to the lowest part of the shell a jet flow is superimposed to the main flow, which is directed towards to the stagnation point and flows tangentially along the shell.The heated jet experiment conducted in the THEADES loop of the KALLA laboratory is nearly 1:1 representation of the lower part of the MEGAPIE target. It is aimed to study the cooling capability of this specific geometry in dependence on the flow rate ratio (Q_main/Q_jet) of the main flow (Q_main) to the jet flow (Q_jet). Here, a heated jet is injected into a cold main flow at MEGAPIE relevant flow rate ratios. The liquid metal experiment is accompanied by a water experiment in almost the same geometry to study the momentum field as well as a three-dimensional turbulent numerical fluid dynamic simulation (CFD). Besides a detailed study of the envisaged nominal operation of the MEGAPIE target with Q_main/Q_jet = 15 deviations from this mode are investigated in the range from 7.5 ≤ Q_main/Q_jet ≤ 20 in order to give an estimate on the safe operational threshold of the target.The experiment shows that, the flow pattern establishing in this specific design and the turbulence intensity distribution essentially depends on the flow rate ratio (Q_main/Q_jet). All Q_main/Q_jet-ratios investigated exhibit an unstable time dependent behavior. The MEGAPIE design is highly sensitive against changes of this ratio.Mainly three completely different flow patterns were identified. A sufficient cooling of the lower target shell, however, is only ensured if Q_main/Q_jet ≤ 12.5. In this case the jet flow covers the whole lower shell. Although for Q_main/Q_jet ≤ 12.5 the flow is more unstable compared to the other patterns most of the fluctuations close to the centerline are in the high frequency range (>1 Hz), so that they will not lead to severe temperature fluctuations in the lower shell material. In this case the thermal mixing occurs on large scales and is excellent.For flow rate ratios Q_main/Q_jet > 12.5 complex flow patterns consisting of several fluid streaks and vortices were identified. Since in these cases the jet flow does not fully cover the lower shell an adequate cooling of the MEGAPIE target cannot be guaranteed and thus temperatures may appear exceeding material acceptable limits.All conducted experiments show a high sensitivity to asymmetries even far upstream. A comparison of the numerical simulation, which assumed a symmetric flow, with the experimental data was due to the experimentally found asymmetry only partially possible.     

Production threshold impact on a GEANT4 calculation of the power deposition in a fast domain: MEGAPIE spallation target

The calculation time in the Monte Carlo simulations consistently represents an essential issue. It is often very long, and its decrease constitutes a challenge for the simulator. Generally, an MC simulation is qualified as quality or not according to two main criteria: the calculation time and the accuracy of the results. However, in most cases, the optimization of one criterion affects negatively the other. Therefore, a compromise between both of them is always required in this kind of simulation. The present work aims at studying the impact of the production threshold(or cut) of the GEANT4 toolkit on the calculation of the power deposition in the MEGAPIE spallation target.The production threshold of secondaries is a GEANT4 intrinsic parameter. It indicates the limit of energy we can reach in the production of secondary particles. This study has allowed us to make the following conclusions. First,the influence of the cut on the calculation of the deposited power depends on the volume size, its arrangement and the importance of the electromagnetic processes occurring within. Second, the accuracy of the calculations can be acceptable only below a given value of the cut energy.Third, this accuracy remains almost unchangeable from a certain value of the cut. The study has also made it possible to explore the prevalence of certain interactions in the zone of spallation in the MEGAPIE target.     

The MEGAPIE 1 MW target in support to ADS development: status of R&D and design

The MEGAPIE project is aimed at designing, building and operating a liquid metal spallation neutron target as a key experiment on the road to an experimental accelerator driven system and to improve the neutron flux at the PSI spallation source. The design of the target system has been completed. The target configuration and the operating conditions have been defined and the expected performance assessed. A preliminary safety analysis has been performed considering normal, off-normal and accident conditions and a corresponding report has been submitted to the authorities for licensing. The experience gained up to now shows that MEGAPIE may well be the first liquid metal target to be irradiated under high power beam conditions.     

Analysis of radioactivity releases in the MEGAPIE reference accident

A computer model for the MEGAwatt Pilot Target Experiment (MEGAPIE) was developed, verified and applied to the analysis of the reference accident, resulting from the hypothetical loss of a Pb–Bi inventory and subsequent cooling down of the target with free convection of the atmospheric air. The radioactivity releases caused by evaporation of activation products from the lead–bismuth films on the surfaces of the target inner structures are estimated. The maximum radioactivity releases were found to be from evaporation of the mercury and polonium estimated as 7.0 × 10¹¹ and 6.2 × 10⁸ Bq, respectively. Thermal conductivity and radiation heat transfer through the gas gaps were found to be more important mechanism of the target cooling down than the atmospheric air convection. The final conclusion made on the basis of the work is that MEGAPIE in the reference accident meets the 1 mSv criterion.     

Verification of the TRACE code against the MEGAPIE transient data

Megawatt pilot target experiment (MEGAPIE) is an international project aimed at demonstrating the feasibility of a liquid lead–bismuth target for spallation facilities at a maximum beam power level of 1 MW. The thermal-hydraulics data measured during the MEGAPIE experiment was used for the TRACE code qualification for transient analysis of liquid metal cooled systems.     

Octalithium plumbate as breeding blanket ceramic: Neutronic performances,synthesis and partial characterization

A neutronic assessment of the performances of a helium-cooled Li₈PbO₆ breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li₈PbO₆ as breeding material. Furthermore, the synthesis and characterization of Li₈PbO₆ by X-ray phase analysis are also discussed.     

冷中子源中子学计算

冷中子源是将热中子慢化成冷中子的实验装置.以蒙特卡罗的三维输运计算程序为基础,使用基于微观中子核反应截面数据(ENDF-BV)库的连续点截面,将堆芯和冷中子源堆内结构结合在一起,建立起完整的蒙特卡罗程序(MCNP)计算模型,完成了冷中子源中子物理学计算.     

Neutronic design of hydride fueled BWRs

This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal-hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.     

Neutronic analysis of iter cryopump system

The ITER Vacuum Vessel has upper, equatorial and lower port structures. The bottom ports are dedicated to the divertor replacement (five ports) and to vacuum pumping by means of cryopumps (four ports). The latest cryopump port design is more complex as it has a pump with a direct view of the vessel (upper cryopump) and a second pump at the end of a branch port (lower cryopump).3D neutronic analyses have been performed in order to study the radiation conditions in and around the port system. In detail, nuclear heating on the cryopump has been calculated updating previous analysis performed in 2003 [L. Petrizzi, ITER CTA Detailed Neutronic Analyses, Final Report on contract EFDA/01-633 ENEA ref NE-VV-R-001 April 2003. Also included in Nuclear Analyis Report NAR ITER ref document G 73 DDD 2W 0.2 (v2.0) March 2006]. Calculations have been performed by means of MCNP 5 Monte Carlo code supplied with FENDL 2.1 library. In this work a new 40° model of ITER has been used in which full details of the cryopump system and remote handling ports have been included as well as the updated divertor components.The paper will present the neutronics results. They consist of nuclear heating on cryopump components; a map of dpa and helium production is provided as well.Gamma doses after shutdown have been calculated around the port flange to have an idea of the possible dose to which the eventual operator will be subject and to plan adequately manual operations.The cryopump is located at a distance of almost 5 m from the mouth of the divertor port and it is 3 m long. Calculations of such deep penetration problem are very challenging require special variance reduction techniques with Monte Carlo codes in order to use in an efficient way the computer resources. These will be described.     

Neutronic design of Indian LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) TBM to be tested in ITER. Liquid lead lithium along with lithium titanate has been adopted as basic material in Indian TBM for neutron multiplication and tritium breeding. RAFMS is used as the structural material and the first wall is cooled by helium. Li-6 enrichment is taken as 60 and 90% in lithium titanate and lead lithium, respectively. The LLCB TBM design is under progress and two design variants are being considered viz. plate design and tube design. In plate design the lead lithium and lithium titanate zones are arranged alternatively and are parallel to the first wall of TBM. In tube design circular tubes of RAFMS are assumed parallel to first wall and lead lithium flows inside the tubes or outside the tubes and lithium titanate is placed accordingly. For the neutronic design of the LLCB TBM, a detailed 3D neutronic model with “look alike” LLCB TBM in equatorial port in ITER has been constructed. A 3D neutron source has been used for the D-T neutrons emitted by plasma. Neutronic study is carried out using Monte Carlo transport code with FENDL-2.1 library with the following objectives: (1) to examine the profiles of heating and tritium production rates in the LLCB TBM, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanical analysis, and (3) to compare the tritium production capabilities of two variants of the geometries. This paper will present the main findings from this neutronic study.     

Neutronic analysis of the ITER Equatorial Port Plug 1

The ITER Equatorial Port 1 will host the following diagnostic systems: the Radial Neutron Camera (RNC), the High Resolution Neutron Spectrometer (HRNS), the Gamma Ray Spectrometer, the Hard X-Rays Monitor, the Pressure Gauges, the Bolometers, the Equatorial Visible/Infrared Wide Angle Viewing System (WAVS), the Neutron Flux Monitor (NFM), the Motional Stark Effect (MSE) system and the Divertor Impurity Monitor (DIM). These diagnostics are integrated inside the Port Plug, a water-cooled stainless steel support structure, which also includes Diagnostic Shielding Modules, designed to provide enough radiation shielding capabilities, to protect the diagnostic systems and to reduce the dose level in the Port Interspace. A new concept for the design of the Port Plug is under consideration: it is based on the installation of the diagnostics inside vertical drawers, completely independent from each other, that are inserted in the Port Plug structure through guiding rails.The paper presents the results of three-dimensional neutronic analyses performed with MCNP5 Monte Carlo code in support of the Port Plug design and integration. The reference ITER MCNP 40° model “Alite” has been updated including the details of the drawers and three diagnostics. Nuclear heating radial profiles have been produced for different toroidal and poloidal positions to be used as input for thermal and thermo-mechanical analyses. 3-D Neutron flux maps have been calculated in order to assess the effect of radiation streaming through all gaps (between the drawers, around the Port Plug and along the diagnostic penetrations) and to provide an estimate of the shielding effectiveness of the new Port Plug concept.     

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.