Full Power Years Research Articles

Preliminary studies of a new accelerator-driven minor actinide burner in industrial scale

Pursuing high transmutation rate of minor actinide (MA), a preliminary conceptual design of a lead-bismuth (LBE) cooled accelerator-driven system (ADS) is proposed in this study. Parametric studies are performed to optimize the neutronics and thermal–hydraulics performances. The proton energy and axial position of the proton beam impact is investigated to obtain high neutron source efficiency and spallation neutron yield. The influences of MA/Pu mixing ratio and the ratio of pin pitch to diameter (P/D) are also optimized to control the burnup reactivity swing and the minimum coolant velocity for adequate cooling. To reduce the power peak, three kinds of power flattening techniques are adopted and compared. The results show that the inert matrix ratio zone loading method seems more versatile. Based on the analyses, an optimized three zone loading pattern is proposed for the 800MWth subcritical core. The total transmutation rate of MA is 328.8kg per effective full power year. Preliminary safety analyses based on the balance of power method (BOP) are performed and the results show that in the reactivity insertion and beam overpower transients, the core shows inherent safety, but the scram is necessary by cutting off the beam current to protect the core from possible damages caused by the loss of flow.

SLIMM-Scalable LIquid Metal cooled small Modular Reactor: Preliminary design and performance analyses

This paper presents a preliminary design of the Scalable LIquid Metal cooled small Modular (SLIMM) reactor for generating 10–100 MWth. It has an estimated operation life of ∼5.8 full-power year (FPY) at 100 MWth and ∼33.3 FPY at 20 MWth and would be fabricated and sealed in the factory, installed below ground, to avoid direct impact by missiles or aircraft, and mounted onto seismic oscillation bearings to resist earthquakes. In-vessel natural circulation of liquid sodium cools the reactor during nominal operation and after shutdown, with the aid of in-vessel Na/Na heat exchanger (HEX) and a tall chimney, while limiting the sodium exit temperature from the core to ≤ 820 K. The design takes advantage of the high heavy metal ratio, high melting point and high thermal conductivity of the UN fuel, and uses HT-9 stainless steel for the UN fuel rods cladding, the core structure, and the reactor primary and guard vessels. The HT-9 is a strong and high corrosion resistant alloy that is compatible with UN and liquid Na at the SLIMM reactor operating temperatures. The active core is comprised of 36 hexagonal assemblies with scalloped-walls BeO shrouds. Each assembly is loaded with 37 UN fuel rods. The core has a negative temperature reactivity feedback, a hard-fast neuron energy spectrum, and independent reactor emergency shutdown (ESD) and control (RC) systems. In the unlikely event of a malfunction of the in-vessel Na/Na HEX, the large sodium mass in the reactor vessel would effectively store heat, until natural circulation of ambient air on the outer surface of the guard vessel can safely remove the decay heat generated in the reactor core. Variable conductance liquid-metal (LM) heat pipes, along the primary vessel wall, transport thermal energy to thermoelectric modules to generate 10 s of kWe of auxiliary power for maintaining the plant's vital functions during nominal reactor operation and in case of a total loss of conventional onsite and offsite powers.

A neutronic design study of lead-bismuth-cooled small and safe ultra-long-life cores

In this paper, two small ultra-long-life lead-bismuth-cooled cores having 48 and 58 effective full power years (EFPYs) of operation life are neutronically designed, and their core physics characteristics including safety-related parameters are analyzed and inter-compared. Each core’s rate is 110MWt (36MWe), and they use depleted uranium and thorium blankets, respectively, for achieving ultra-long life and high burnup while the ternary metallic fuels having TRU from LWR spent fuels are adopted as driver fuels. They adopt a low power density for ultra-long life and for considering some possibilities of high peak power density. This work provides a consistent comparison of the effects of the depleted uranium and thorium blankets on the performances of small ultra-long-life lead-bismuth-cooled cores. The results of the detailed neutronic analyses show that the core with depleted uranium has a significantly longer life because of the better breeding characteristics of depleted uranium versus the core with a thorium blanket and that all of the cores have negative reactivity coefficients except for the very small positive coolant expansion reactivity coefficients and low peak linear power densities. Additionally, a detailed decomposition analysis is performed to better understand the effects of coolant voiding on reactivity.

Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated pressure tube which has been removed from Kakrapar Atomic Power Station-2 after 12.76 Full Power Years of operation for post irradiation examination studies at Bhabha Atomic Research Centre. This paper highlights the basic theory of measurement, results from the qualification trials on Zr–2.5Nb material having different mechanical properties and reactor operated irradiated pressure tube.

Neutronics analysis for ITER cable looms

This paper briefly presents the results of neutronics analysis performed in support for the design development of the ITER diagnostics cabling system including its cable looms. The cable looms are distributed all around the ITER device and attached to the inner wall of the Vacuum Vessel (VV) and under the divertor cassettes. The results include distributions of neutronics characteristics calculated with the MCNP5 3D Monte Carlo code assuming ITER operation with 0.54 Full Power Years on 500MW fusion power of DT neutron source defined in the standard ITER B-lite neutronics model. Neutron and photon fluxes, neutron energy group fluences, nuclear heating, neutron damage, and neutron-induced transmutation of the cable looms’ materials have been analyzed in this paper. This work provides data for more comprehensive integral analysis of the cabling system with consideration of thermo-mechanical, electromagnetic, and structural aspects. The temperature gradients along the cable looms and nuclear transmutation of copper could result in formation of thermocouples due to the Radiation-Induced Thermoelectric Sensitivity (RITES) effect, which should be reduced in order to enhance the performance of the magnetic sensors. Therefore, reliable thermal contact should be established in the places of attachments of cable looms to the VV inner wall and the divertor cassette body.

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO2–ThO2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO2–ThO2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO2–UO2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.

Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200°C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500°C and should be lower than 30% of the total amount produced over a reactor operating time corresponding to about 11 effective full-power years. Moreover, it would mainly concern the tritium located close to the free surfaces. Furthermore, the total extraction of the remaining tritium in graphite waste should be more efficient in dry inert gas than in humid gas, but would require temperatures higher than 1300°C for the total removal of the most deeply located deuterium.

Pre-conceptual study of small modular PbBi-cooled nitride fuel reactor core characteristics

In this paper a pre-conceptual neutronics study on a small modular Pb-Bi cooled reactor with nitride fuel (SMoPN) is presented. The SMoPN is designed to meet the requirements for nuclear energy expansion in the next decades, by using the plutonium and thorium nitride fuel to increase the efficiency and performance of fuel. Based on the existing experiences of nuclear reactor, the primary design parameters were provided to match the design goals by the whole core three-dimensional calculation. The nitride fuel, stainless steel cladding and Pb-Bi coolant have a perfect compatibility with each other, as well as with excellent neutronics characteristics. High conversion ratios have been achieved to ensure 20 effective full power years (EFPYs) without refueling and shuffling. During the core lifetime, the burn-up swings slightly due to the perfect breeding capability. All of the important reactivity coefficients are negative to assure the SMoPN holding passive safety. The control system can provide enough shutdown margins in both normal and abnormal operational conditions. Therefore, the SMoPN concept satisfies completely the advanced design idea and the requirements of advanced nuclear reactor system.

Fracture toughness of irradiated Zr–2.5Nb pressure tube from Indian PHWR

Fracture toughness of irradiated Zr–2.5Nb alloy pressure tube, fabricated by the cold pilgering and stress relieving route, was evaluated using disk compact tension type specimens. These specimens were punched out from the irradiated pressure tube (S-07), which was in service for about 8 effective full power years of reactor operation in the Kakrapar Atomic Power Station-2 (KAPS-2). The tests were carried out remotely inside a lead shielded enclosure. Crack growth during the test was measured using the direct current potential drop technique. The irradiated pressure tube showed low fracture toughness at 25°C. The fracture toughness increased with increase in temperature up to 250°C but was practically unaffected with further increase in temperature up to 300°C. This paper discusses the fracture behavior of irradiated Indian pressure tube material and compares it with other data available.

Comparative CFD analyses of liquid metal cooled reactor for lunar surface power

This paper presents the results of comparative CFD and thermal-hydraulics analyses of the Solid Core–Sectored Compact Reactor (SC-SCoRe) for lunar surface power. This fast-neutron spectrum, liquid NaK-56 cooled reactor is loaded with highly enriched UN fuel. It nominally generates 1.0MWth for ∼21 full-power years at NaK-56 inlet and exit temperatures of 850K and 900K. The analyses examine the realizable two-layer k–ɛ and the Shear Stress Transport (SST) k–ω turbulence models, with different numerical mesh refinements, for simulating the performance of the SC-SCoRe core during nominal operation as well as investigate the effect on the computation time and calculated parameters. In addition, the results calculated for a single tri-lobe flow channel are compared to those obtained using the Detached Eddy Simulation (DES), a hybrid LES and RANS turbulence model. The numerical mesh refinement beyond 2×107 cells in the flow channels and the choice of turbulence model slightly affect the calculated fuel, core structure, and liquid metal temperatures. They more strongly affect the pressure losses and the intensity of flow mixing and the formation of turbulence eddies in the inlet and exit plenums and exit duct. With the same numerical mesh refinements, the total computation time with the k–ω model is 40–120% longer than with the k-ɛ model, while the calculated operation parameters are almost identical. The flow field provided by the k–ω model offers finer details of the flow mixing and the eddy vortices forming in the upper plenum and the exit duct. Comparisons of the heat transfer and pressure loss results of the DES model to those of the k–ɛ and k–ω turbulence models for a single tri-lobe, show the calculated temperatures, velocity profiles, and pressure losses differ only slightly. The DES model results of the Nusselt number along the channel are consistently higher than those calculated using the k–ω and the k–ɛ models. The time to complete the calculations with the DES model, on the same hardware, was 38 and 86 times that with the k–ω and the k–ɛ models, respectively. The results for NaK-56 liquid flow in a circular tube with uniformly heated wall show that the Nu values using the k–ω and k–ɛ turbulence models are consistent with the reported empirical correlations.

Development of rolled joint detachment system for low enriched uranium based advanced heavy water reactor

Advanced heavy water reactor-low enriched uranium (AHWR-LEU) is designed for high discharge burnup and large scale commercial utilization of thorium with LEU as an external feed in a once through fuel cycle. It is a 300MWe vertical pressure tube type, thorium based reactor, cooled by boiling light water and moderated by heavy water. The reactor incorporates a number of passive safety features and is associated with a fuel cycle having reduced environmental impact and offers more proliferation resistance compared to contemporary light water reactors (LWRs). Some distinguishing features which make AHWR-LEU unique from other conventional nuclear power reactors are core heat removal by natural circulation, negative void coefficient of reactivity under all event scenarios, easily replaceable pressure tubes, long design life, passive safety features, etc. The long design life of the reactor is possible to be met only when the pressure tube which has a limited life of 15–20 full power years (FPYs) due to a life limiting mechanism called irradiation enhanced creep and growth is easily replaceable. Pressure tube is joined at its ends to SS 403 end fittings through rolled joints. A system has been designed and developed for detachment of pressure tubes from the end fitting and its associated hardware without disturbing the surrounding components or coolant channels so that the end fitting can be reused. The system works on the principle of shock heating and fast cooling of the pressure tube in the rolled joint region and application of axial load. Though rolled joint detachment by induction heating has been used earlier in CANDU reactors for removal of pressure tubes and calandria tubes, the present design incorporates a liquid nitrogen based faster cooling arrangement and an integrated load application and gripping module for detachment and removal of pressure tube from the end fitting in much lesser time. This paper describes the methodology of rolled joint detachment, challenges involved in the design, features of the detachment system, findings of experimental trials and observations.

The dynomak: An advanced spheromak reactor concept with imposed-dynamo current drive and next-generation nuclear power technologies

A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000MW from a thermal output of 2486MW was achieved, yielding an overall plant efficiency of approximately 40%.

TEM characterization of in-reactor neutron irradiated CANDU spacer material Inconel X-750

The irradiation induced defects in CANDU Inconel X-750 spacers, which were removed from reactors after about 14 effective full power years, were examined by transmission electron microscopy (TEM). The spacers in the form of garter springs were reported to operate at various temperatures depending on locations. Two samples from different locations with different estimated irradiation temperatures were tested: (1) ∼180°C at 6 o’clock position and (2) ⩾300°C at 12 o’clock position. Obvious temperature effects were observed. In the ∼180°C irradiated sample, a high density of small lattice defects (1–3nm) developed during irradiation, including stacking fault tetrahedra and both 1/3 〈111〉 and ½ 〈110〉 type dislocation loops. A uniform distribution of small cavities (∼1–3nm) was observed. In >300°C irradiated sample, apart from small point defect clusters, large Frank type interstitial loops presented. The sizes of the cavities were also greater than those in the ∼180°C irradiated sample. The distribution of cavities was more heterogeneous and an obvious agglomeration of cavities to grain boundaries and phase boundaries were observed. In both samples, dissolution of the primary strengthening phase γ′ was noted.

Burnup characteristic of CERMET and TRISO fuel particles used in Boiling Water Reactor with Pebble-Bed Coated Particles

The reactor concept of boiling water reactor with pebble-bed coated particles (BWR-PB), as a 890MW(th) without the need of on-site refueling, is originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by boiling light water. At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating outer layer during a fuel cycle. This integrity problem, under low temperature irradiation, related to the accumulation of atomic displacements in the graphite lattice structure (Winger Energy) is due to insufficient annealing at temperatures below 260C. Instead, an alternative Spherical Cermet Fuel (SCF) concept is proposed. It consists of coated UO2 kernels embedded in a Zirconium matrix which is then over-coated with a protective outer layer of Zirconium. This new SCF concept may offer the following advantages: high thermal conductivity, low fuel temperatures, good fission product retention and good performance at high burnup. The SCF concept aims at long life operation capability and neutronic feasibility in BWR-PB reactor. This paper is focused on the evaluation of some nuclear parameters of a SCF concept, for a BWR-PB reactor, and the comparison with the previous TRISO fuel calculation results. The calculated burnup characteristic parameters are k-infinity, Instantaneous conversion ratio, Fuel isotopic and Neutron energy spectra. The investigation of this latter has found out that burnup characteristic parameters are changed slightly between the two fuel concepts. In fact, using SCF reduces the lifetime operation by 2 equivalent full power years in each BWR-PB assembly’s part, mainly due to the larger thermal component in the neutron spectrum for TRISO fuelling compared to the SCF one. The concentration of Uranium in the SCF spent fuel is similar to that contained in TRISO. However, the plutonium isotopic composition shows a significant deviation between the two concepts. Besides, in both fuel concepts, the fuel fertile matter is more exploited from the bottom to the top of BWR-PB fuel assembly.

Conceptual design study of Hyb-WT as fusion–fission hybrid reactor for waste transmutation

This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60w/o, Zr 40w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle.

Predicting tube repair at French nuclear steam generators using statistical modeling

Electricité de France (EDF) currently operates a total of 58 Nuclear Pressurized Water Reactors (PWR) which are composed of 34 units of 900MWe, 20 units of 1300MWe and 4 units of 1450MWe. This report provides an overall status of SG tube bundles on the 1300MWe units. These units are 4 loop reactors using the AREVA 68/19 type SG model which are equipped either with Alloy 600 thermally treated (TT) tubes or Alloy 690 TT tubes. As of 2011, the effective full power years of operation (EFPY) ranges from 13 to 20 and during this time, the main degradation mechanisms observed on SG tubes are primary water stress corrosion cracking (PWSCC) and wear at anti-vibration bars (AVB) level.Statistical models have been developed for each type of degradation in order to predict the growth rate and number of affected tubes. Additional plugging is also performed to prevent other degradations such as tube wear due to foreign objects or high-cycle flow-induced fatigue. The contribution of these degradation mechanisms on the rate of tube plugging is described.The results from the statistical models are then used in predicting the long-term life of the steam generators and therefore providing a useful tool toward their effective life management and possible replacement.

Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S&B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20MWth. A S&B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S&B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S&B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S&B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO2 fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S&B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

Reactor pressure vessel integrity assessment by probabilistic fracture mechanics – A plant specific analysis

The probabilistic fracture mechanics method is widely adopted in nuclear power plant industry, especially for the structural integrity assessment of reactor pressure vessels. In this work, the plant specific analyses of Taiwan's boiling water reactors, BWR/4 and BWR/6, under the design transients were performed by the FAVOR (Fracture Analysis of Vessel – Oak Ridge) code. The difference of reactor pressure vessel geometry, alloying elements, neutron fluence, and loading conditions were taken into account in the probabilistic fracture mechanics analyses. The failure probabilities of the axial welds at 64 effective full-power year (EFPY) for the analyzed BWR/4 and BWR/6 are less than 2.8 × 10−10/year and 1.5 × 10−4/year, respectively. Furthermore, the circumferential welds present much smaller failure probabilities. The results of this work are useful for the subsequent aging management of the operating reactor pressure vessels.

SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

The effect on radiation damage of structural material in a hybrid system by using a Monte Carlo radiation transport code

In this study, the molten salt-heavy metal mixtures 99–95% Li20Sn80-1-5% SFG-Pu, 99–95% Li20Sn80-1-5% SFG-PuF4, 99-95% Li20Sn80-1-5% SFG-PuO2 were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. 9Cr2WVTa ferritic steel with the width of 4cm was used as the structural material. The parameters of radiation damage are proton, deuterium, tritium, He-3 and He-4 gas production rates. In this study, the effects of the selected fluid on the radiation damage, in terms of individual as well as total isotopes in the structural material, were investigated for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent version of the MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.