Mitigation technologies for damage induced by pressure waves in high-power mercury spallation neutron sources (IV) – measurement of pressure wave response and microbubble effect on mitigation in mercury target at J-PARC –

A mercury target system to produce neutron beams has been operated at the spallation neutron source in the Japan Proton Accelerator Research Complex (J-PARC). Pressure waves are generated in mercury by rapid heat generation due to bombardment by high-intensity short-pulse proton beams. The pressure waves not only cause cyclic stress but also induce the cavitation damage on the target vessel. Reduction of these pressure waves is important from the viewpoint of extending the lifetime of the target vessel in future power-up operations. The injection of microbubbles into mercury is effective for reducing pressure waves. Accordingly, a microbubble generator was installed in the mercury target vessel and an in situ diagnostic system that measures the displacement velocity of the target vessel induced by the pressure waves was also set up in J-PARC to investigate the effect of proton beam condition and the effect of the microbubbles. Consequently, we found that the peak displacement velocity of the target vessel decreased owing to microbubble injection. The ratios of the peaks obtained with bubble injection to that without bubble injection were 1/3 and 2/3 when the injected gas fractions were 0.4% and 0.1%, respectively.     

Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water reactor (BWR). Therefore, the evaluation of SCC propagation is important for the safe maintenance of the core shroud. This paper attempts to predict the remaining life of the core shroud due to SCC failures in BWR conditions via SCC propagation time calculations. First, a two-dimensional finite element method model containing H6a girth weld in the core shroud was constructed, and the weld processing was simulated to determine the weld's residual stress distribution. Second, using a basic weld residual stress field, the SCC propagation was simulated using a node release option and the stress redistribution was calculated. Combined with the J-integral method, the stress intensity factors were calculated at depths of 2, 3, 4, 8, 12, 16, 19, 22, 25 and 30 mm in the crack setting inside the core shroud; then, the SCC propagation rates were determined using the relation between the SCC propagation rate and the stress intensity factor. The calculations show that the core shroud could safely remain in service after 9.29 years even when a 1-mm-deep SCC has been detected.     

Damage Diagnostic of Localized Impact Erosion by Measuring Acoustic Vibration

High power spallation targets for neutron sources are being developed in the world. Mercury target will be installed at the material and life science facility in J-PARC, which will promote innovative science. The mercury target is subject to the pressure wave caused by the proton bombarding mercury. The pressure wave propagation induces the cavitation in mercury that imposes localized impact erosion damage on the target vessel. The impact erosion is a critical issue to decide the lifetime of the target. The electric Magnetic IMpact Testing Machine, MIMTM, was developed to produce the localized impact erosion damage and evaluate the damage formation. Acoustic vibration measurement was carried out to investigate the correlation between the erosion damage and the damage potential derived from acoustic vibration. It was confirmed that the damage potential related with acoustic vibration is useful to predict the damage due to the localized impact erosion and to diagnose the structural integrity.     

Technical investigation on small water leakage incident occurrence in mercury target of J-PARC

Japan Proton Accelerator Research Complex experienced failures of two mercury targets, which were Target #5 and #7, in 2015 when the facility was operating with a proton beam power of 500 kW. The failures involved coolant water leak from the water shroud. In this paper, we investigate the root cause of the Target #5 failure. The results of the visual inspections, mockup tests, and analytical evaluations suggested that the water leak was caused by the possible combination of two incidents. One was the diffusion bonding failure due to the large thermal stress induced by welding of the bolt head during the fabrication process, and the other was the thermal fatigue failure of the seal weld due to the repetitive beam shutdown during beam operation. Though the investigation into the root cause of the Target #7 failure is still going on, these target failures point to the importance of eliminating initial defects and the need to secure the rigidity and stability of welded structures. The next mercury target, Target #8, was fabricated with an improved design and fabrication process to reduce the possibility of similar failures. The beam operation of this mercury target is planned to be started in October 2017.     

Dynamic Response of Mercury Subjected to Pressure Wave

The Japan Atomic Energy Agency (JAEA) is carrying out R&D for constructing the facility of high intensity spallation neutron source which may bring us innovative science fields. A high power pulsed proton beam will be injected into a mercury target for nuclear spallation reaction. Due to the pulsed proton injection, mercury is heated rapidly and pressure waves are generated. The mercury target vessel, in which mercury is enclosed, is subjected to the pressure waves. Dynamic response of mercury, such as the pressure waves in mercury, is important to evaluate the integrity of the mercury target vessel. In order to examine the dynamic response of mercury, we have carried out impact experiments on mercury by using the split-Hopkinson pressure-bar (SHPB) technique. Numerical analyses were also carried out to verify the analytical model by using an explicit FEM code. It is found that the analytical results approximately represented the experimental results in a very early stage of impact. And it is recognized that the stiffness of mercury under impact condition was independent of the impact velocity in this experimental range. Furthermore, many pits were found on the surfaces being in contact with mercury.     

Extension of Spallation-Based BNCT Concept to Medium- to High-Energy Accelerators

Following our systematic studies on a cyclotron-based neutron source for Boron Neuron Capture Therapy (BNCT), we expanded our investigation to include medium- to high-energy accelerators with a proton energy range of 30 to 600MeV using a spallation reaction and appropriate moderating materials. The results showed that a spallation-based neutron source for BNCT can be realized without any significant engineering difficulties by applying a commercially available high-current cyclotron and the linear accelerator at the Japan Proton Accelerator Research Complex (J-PARC). A proton therapy accelerator cannot be directly applied for BNCT due to the low beam current. However, the use of a proton therapy accelerator will be especially effective because the proton therapy for localized cancers is complemented with BNCT, which is ideal for treating nonlocalized and radio-resistant cancers.     

Water flow experiments and analyses on the cross-flow type mercury target model with the flow guide plates

A mercury target is used in the spallation neutron source driven by a high-intensity proton accelerator. In this study, the effectiveness of the cross-flow type mercury target structure was evaluated experimentally and analytically. Prior to the experiment, the mercury flow field and the temperature distribution in the target container were analyzed assuming a proton beam energy and power of 1.5 GeV and 5 MW, respectively, and the feasibility of the cross-flow type target was evaluated. Then the average water flow velocity field in the target mock-up model, which was fabricated from Plexiglass for a water experiment, was measured at room temperature using the PIV technique. Water flow analyses were conducted and the analytical results were compared with the experimental results. The experimental results showed that the cross-flow could be realized in most of the proton beam path area and the analytical result of the water flow velocity field showed good correspondence to the experimental results in the case when the Reynolds number was more than 4.83×10⁵ at the model inlet. With these results, the effectiveness of the cross-flow type mercury target structure and the present analysis code system was demonstrated.     

Prediction of residual stress distributions due to surface machining and welding and crack growth simulation under residual stress distribution

In nuclear power plants, stress corrosion cracking (SCC) has been observed near the weld zone of the core shroud and primary loop recirculation (PLR) pipes made of low-carbon austenitic stainless steel Type 316L. The joining process of pipes usually includes surface machining and welding. Both processes induce residual stresses, and residual stresses are thus important factors in the occurrence and propagation of SCC. In this study, the finite element method (FEM) was used to estimate residual stress distributions generated by butt welding and surface machining. The thermoelastic-plastic analysis was performed for the welding simulation, and the thermo-mechanical coupled analysis based on the Johnson-Cook material model was performed for the surface machining simulation. In addition, a crack growth analysis based on the stress intensity factor (SIF) calculation was performed using the calculated residual stress distributions that are generated by welding and surface machining. The surface machining analysis showed that tensile residual stress due to surface machining only exists approximately 0.2 mm from the machined surface, and the surface residual stress increases with cutting speed. The crack growth analysis showed that the crack depth is affected by both surface machining and welding, and the crack length is more affected by surface machining than by welding.     

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.     

Present status of study on development of materials resistant to radiation and beam impact

Pulsed spallation neutron sources for the materials structure science are severely influenced by beam impact and radiation damage. We have developed the materials strong to these influence since 2004. In this paper, recent topics are described concerning the development of intergranular corrosion (IGC)-resistant austenitic stainless steel for target vessel and window, radiation-resistant ultra-fine grained tungsten materials (W-TiC) for a solid target, CrN film on a tungsten target by means of a molten-salt method, surface treatment of stainless steel for pitting damage in mercury target. Bubble behavior at the interface of mercury and window glass was also observed to clarify the phenomenon of the pitting damage.     

Pressure wave propagation in an LWR downcomer and resultant load on the core shroud due to rupture of a primary coolant loop

Pressure differences and the resultant dynamic load act on the core shroud when pressure waves propagate in the downcomer of a light water reactor (LWR) pressure vessel after rupture of the primary pipe has occurred. An equivalent geometry, i.e. a diverging duct is used to solve by Euler and wave equation for acceleration and velocity of the fluid behind the wave front, that the two-dimensional, time-dependent pressure distribution, induced by the wave propagation, can be calculated. The assumptions lead to an approximate but conservative value of the resultant core shroud load.     

Probabilistic analysis of core shroud cracks

Traditional limit load analysis and fracture mechanics analysis have been applied to evaluate the integrity of the degraded nuclear power plant components. Although these methodologies are generally accepted by the regulatory authorities in the nuclear industry, conservatism introduced by the uncertainties of inspection, material property, crack geometry, applied loading, neutron environment, etc. is recognized to have great impact on the evaluation accuracy. A probabilistic analysis may overcome this shortcoming and reveal some additional insight to the problem. The purpose of the present study is to apply probabilistic methods to analyze the degraded core shroud, and to predict the quantitative risk of the cracked shroud. In the analysis, the loading condition, crack growth rate, material properties and existing defects are all considered random. A sample analytical result shows that, based on some previously observed data and under certain assumptions, the crack-through probability of the studied core shroud is in the order of 10⁻⁷ after 13 cycles of operation. The probability will increase considerably through operation cycles or operation years if no repair action is taken.     

Effects of thermal load and cooling condition on weld residual stress in a core shroud with numerical simulation

Stress corrosion cracking (SCC) in the heat affected zone is the primary damage form due to weld residual stress, corrosion and neutron irradiation environment in the core shroud of a boiling water reactor. The distribution of weld residual stress around a weld is necessary to be clarified to evaluate the structural integrity of core shroud for SCC. Moreover, studying the effects of welding parameters on residual stress on reducing the residual stress is very important to suppress the initiation and propagation of SCC.In this paper, we used a finite element method (FEM) to clarify the distribution of weld residual stress around the sixth horizontal weld (H6a) between the lower ring and the cylinder in the core shroud. The simulation results of axial stress were consistent with the experimental results at the inside and outside surfaces of the core shroud, respectively. The effects of thermal loads and cooling conditions were also investigated with the same model. We simulated the welding progress with water cooling on the inside and outside surfaces of the core shroud in order to study the influence of cooling conditions on the residual axial stress around the weld. The simulation results indicated that water cooling decreased the residual axial stress at the same side due to changing the temperature-affected fields. Moreover, with fixing the peak temperatures of weld passes, the simulation results of the distribution of residual axial stress by the thermal loads with different heating time were compared. The simulation results suggested that the heating time was expected to be longer and the heat flux to be smaller for reaching the small tension residual axial stress or even compression stress around the H6a weld.     

Measurements of neutronic characteristics of rectangular and cylindrical coupled hydrogen moderators

Extensive simulation calculations were performed in the design studies of the coupled hydrogen moderator for the pulsed spallation neutron source of the Japan Proton Accelerator Research Facility (J-PARC). It was indicated that a para-hydrogen moderator had an intensity-enhanced region at the fringe part, and that pulse shapes emitted from a cylindrical para-hydrogen moderator gave higher pulse-peak intensities with narrower pulse widths than those from a rectangular one without penalizing the time-integrated intensities. To validate the peculiar distribution and advantages in pulse shapes experimentally, some measurements were performed at the neutron source of the Hokkaido University electron linear accelerator facility. It was observed that the neutron intensity was enhanced at edges of the para-hydrogen moderators, whereas it decreased at the same part of the ortho-rich-hydrogen moderator, where the dimension of those moderators was 50 mm in thickness and 120 mm in width and height. The spatial distribution and pulse shapes were also measured for a cylindrical coupled para-hydrogen moderator that has the same dimensions as for the coupled moderator employed for J-PARC. The measured results from the cylindrical moderator were consistent with the results obtained in the design studies for the moderator for J-PARC.     

Computational Fluid Dynamic Studies of the MEGAPIE Spallation Source Target and Safety Vessel

In this paper, we describe some numerical investigations that have been undertaken in demonstrating the cooling principle of the MEGAPIE liquid-metal neutron spallation source target using computational fluid dynamics (CFD). Simultaneously, stresses in the structural components have been examined using finite element method (FEM) techniques, with an in-house CFD/FEM interface program employed to ensure full consistency of data at the local level. Results for steady-state operation of the target show that the critical lower target components are adequately cooled under normal operating conditions, and that stresses and displacements are well within tolerances. With unexpected overfocusing of the beam, there is the potential for structure failure of the target window. Detailed analysis has shown that 35% overfocusing canbe tolerated, and that even if the target window is breached, the D2O-cooled safety vessel positioned around the target will remain intact, even with 100% beam overfocusing. Transient analysis of a thermal shock incident is also described in which a jet of cold D²O is imagined to impinge on the target window following rupture of the nearby water circuit. Results indicate that the margin of error to ductile-brittle transition is large enough for the integrity of the window not to be at risk from this incident.     

Development of remote handling technology of liquid lithium target and replaceable back plate with lip seal in IFMIF-EVEDA

The IFMIF is an accelerator-based intense neutron source for testing candidate fusion materials. Intense neutrons equivalent to neutron irradiation damage of about 50 dPa/y are emitted inside the Li flow through a back plate. Around the back plate, a lip seal made of 316 L is welded by laser-welding system for replacement by remote handling. The back plate will be designed for replacement at least every year. According to material tests of the lip seal weld joint (316 L/316 L) at room temperature, significant deterioration was not observed. Further investigation of the welding process of the lip seal such as a welding direction and a welding joint shape is in progress. Remote handling procedure of the back plate is examined. At first, three lip seal joints of connection piping will be cut by the laser cutting/welding device and then the target assembly with the back plate will be moved to a hot cell. The back plate lip seal will be cut by the laser arm in the hot cell. After machining and Li cleaning of the lip seal, a new back plate will be welded and moved to test cell/target room.     

压水堆重反射层堆芯核热耦合高精度计算分析研究

重反射层的应用可提高反应堆中子经济性,其结构和中子吸收特性均与压水堆常规围板/反射层差异较大,因此对核设计程序的计算分析能力提出了新的要求。为分析重反射层建模方案对堆芯中子学计算结果的影响,使用先进中子学程序SCAP N和确定论堆芯高保真模拟程序NECP X对压水堆重反射层问题进行了高保真模拟,分析了5种反射层建模方案下计算结果的差异,并将高精度计算结果与商用核设计程序系统进行了对比。数值结果表明,重反射层水洞内冷却剂温度变化对计算结果影响较小;相较精确建模方案,重反射层铁水打混建模方案造成的反应性计算偏差在±30 pcm以内、组件相对功率分布计算偏差在±2%以内。     

Use of low neutron leakage fuel loads to decrease the radiation load on a VVÉR-1000 vessel

It is shown on the basis of data obtained at Ukrainian nuclear power plants that fuel loads with low neutron leakage can be used effectively to decrease the radiation load on the reactor vessel. The characteristics of 104 fuel loads and the results of a determination of the radiation load on the vessel are analyzed to develop a criterion according to which a VVéR-1000 fuel load can be classified as a load with low neutron leakage. It is shown that the following condition can be chosen as such a criterion: the run-averaged relative power release in all protruding fuel assemblies must be less than 0.57. Different variants of the arrangement of the VVéR-1000 core are examined and analyzed. It is shown that placing burned-out fuel assemblies along the periphery of the core and decreasing the number of neutrons leaving the core do not always result in a lower neutron load on the reactor vessel. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 93–97, August, 2006.     

Sustained load crack growth in A533B-1 steel under neutron irradiation in a water environment

A study of the combined effects of radiation, water and temperature on sustained load crack growth behavior of reactor pressure vessel steel A533B-1 is reported. To complete this study wedge opening loading (WOL) T-type fracture toughness specimens were prepared from a sample of A533B-1 steel which had a copper content of 0.13%. The crack length change was measured after 939 hr of irradiation in a water environment. An electrical potential method was successfully used to measure the crack length of rusted radioactive specimens. Sustained load crack growth occurred at initial stress intensity factor K_Ii as low as . The value of stress corrosion cracking threshold factor K_Iscc after neutron irradiation in a water environment appears to be in the range of . The results of neutron irradiation in a water environment are to apparently increase the susceptibility of A533B-1 steel to stress corrosion cracking and hydrogen embrittlement.     

Neutronics and photonics of inertial confinement fusion pellets with internal breeding

The neutronics and photonics performance of a pellet with a small DT core spark trigger, surrounded by a large volume of D to enable tritium and He-3 breeding, is examined. The response to a 70% DD and 30% DT composite neutron spectrum is calculated using either W, Be, or Pb as structural materials at core density radius products ranging from 9.42 to 94.2 kg/m². At a core density-product of 94.2, the DT neutron source leads to an excess particle multiplication of 0.43 neutrons per source neutron. The percentage of energy leakage from the pellet in the form of escaped neutrons is 42.3% of the source energy for the DT source, and 28.8% for the DD source. The gamma-ray energy percentage deposited in the pellet is 26.7% for the DT source and 106.6% for the DD source. For the pellet with the composite source, the energy multiplication factor is 1.27. Thus the large DD contribution to the composite neutron source results in the pellet performing many of the functions normally reserved for the blanket such as spectral softening, breeding, and neutron and energy multiplication. The neutron energy leakage is 38.4% of the source energy for the composite source. It is estimated that the neutron energy leakage amounts to 10% of the fusion energy, compared with 70% as neutron energy in a DT pellet. These results are significantly different from those encountered in conventional DT inertial confinement designs, and thus lower tritium inventories, higher power densities, reduced radiation damage, and materials activation of the reactor coolant and structure may be achievable.