Equivalent Full Power Research Articles

Evaluation of 99Mo production in a small modular thorium based molten salt reactor

Producing medical radionuclide 99Mo in molten salt reactors is a very attractive choice to solve its global shortage. In this study, we evaluated the yield of 99Mo in the small modular thorium based molten salt reactor (SM-MSR). Firstly, the fuel burn-up analysis of SM-MSR was carried out by an in-house developed code (MOBAT), which takes the unique characteristics of molten salt reactor into account, and the variation of fission yield of 99Mo with burn-up time was obtained. The minimum value of the fission yield of 99Mo is 1.13 × 10−3 (6-day TBq/MW/s) at approximately 600 equivalent full power day. Then, based on the behavior of noble metals in the fuel salt and the experimental results of MSRE gas sample composition measurements, 99Mo migration probability from the primary loop to the off-gas module with the burn-up time was calculated, and the equilibrium value of the migration probability is obtained as 18.4%. When the load factor of SM-MSR is 0.75 (300 MWth), the annual 99Mo amount of migration to the off-gas module would be 1.96 × 106 (6-day TBq) under the most conservative calculation. Finally, a filter system was added in the off gas module, and 99Mo would be extracted from the off gas. As long as the utilization percentage of 99Mo in the off gas module can reach 0.94%, the global demand under current data could be met, which implies a huge additional economic value for the SM-MSR.

Neutronics Analysis of Small Compact Prismatic Nuclear Reactors for Space Crafts

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235U mass and the total reactor mass.

Comparative analyses of coated and composite UN fuel – Monte Carlo based full core LWR study

Longer continuous operation of a nuclear reactor leads to higher availability of nuclear power plant, entailing economic gain. In this context the option being explored, recently, is the use of higher density fuels as compared to current fuels i.e. UO2. Uranium mono nitride (UN) fuel is one of the options being explored in this regard. UN fuel has been used in nuclear industry for a long time with fast reactor option. More recently, studies have targeted its use in Light Water Reactor (LWR) environment. The main problem with using UN fuel in LWR is its potential reaction with water which produces hydrogen. Researchers have proposed different approaches to overcome this problem.One option is the use of coatings around UN fuel pellet to avoid the direct contact of water with fuel. The second option is use of a secondary phase (10 vol. percent) like ZrO2 which can make an oxide layer in case of contact with water and protect the main phase, Uranium mono nitride (80 vol. percent). Remaining 10 vol. percent is considered to be consumed by porosity. This study aims at comparison of neutron physics behavior of both these options in LWR environment. The upshot of the study is to quantify the impact of adding layers or secondary phase with respect to pure/complete UN fuel. To study these effects, two theoretical densities i.e. 95% and 80% for UN fuel are chosen for analyses. To avoid the problem of C-14 production from N-14, fuels studied are considered to be having 100% N-15.A validated model of Benchmark for Evaluation and Validation of Reactor Studies (BEAVRS) is used to perform all the analyses. Integral neutron physics parameters like neutron energy spectra, Radial Peaking Factors, Axial Peaking Factor, Doppler coefficient of reactivity, Isothermal Coefficient of reactivity and Temperature defect, Control Rod worth and excess reactivity for whole core are compared at Beginning of Cycle (BoC). Burnup obtained by different fuel option is also compared. Consistent with pin-cell based earlier findings, this full 3D analyses with UN based fuel shows noticeable spectral hardening leading to decrease in the value of control rod worth and less negative Doppler coefficient of reactivity while power peaking factors remain mostly unchanged. The economic advantage of switching to UN based fuels is expected when UN fuel above 80% TD is used. Approximately 19% increase in Equivalent Full Power Days (EFPDs) is witnessed by using 95% TD UN based fuels.

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14MW. ORPHEE core is 90cm height and has a cross section of 27×27cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® (Both and Peneliau, 1996; Trama et al., 2012; Brun, 2011). A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 (Sanchez, 2009). The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000MWth output and tritium self-sufficiency (TBR>1.15), while the required maximum fusion powers are lower than 500MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

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.

Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

The Atoms For Peace Reactor (AFPR-100), as a 100MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY.

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.

Submersion-Subcritical Safe Space (S 4) reactor

The Submersion-Subcritical Safe Space (S 4) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo–14% Re solid core, loaded with uranium nitride fuel, cooled by He–30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively.

An additional performance of HTRS: the waste radiotoxicity minimisation

The management of radioactive waste is a key issue for the present and future use of nuclear energy. In this frame, high temperature reactors (HTRs) have, among others, the capability to burn actinides. After a short introduction on HTRs, the performances of two MC-based burnup codes (Monte Carlo continuous energy burnup and MONTEBURNS) in assessing the ability of these reactors to burn actinides are compared. These codes are necessary for performing ultra-high burnup calculations on HTRs. The best one, in this specific case, results to be MONTEBURNS. It was analysed using HTRs loaded with the following: (1) 1st generation Pu, 600 equivalent full power days; (2) 2nd generation Pu, 645 equivalent full power days; and (iii) 33% 1st generation Pu and 67% Th, 705 equivalent full power days. Finally, it is possible to conclude that HTRs can reduce time when the waste is considered dangerous. Even if the amount of reduction does not solve the whole problem, it represents an important step in the management of radioactive waste.

Steam Generator Materials Performance in High Temperature Gas-Cooled Reactors

High temperature gas-cooled reactor (HTGR) systems feature a graphite-moderated, uranium-thorium, all-ceramic core and utilize high pressure helium as the primary coolant. The steam generators in these systems are exposed to gas-side temperatures approaching 760°C (1400°F) and produce superheated steam at 538°C (1000°F) and 16.5 MPa (2400 psi). Thus, the design and development of steam generators for these systems require consideration of time-dependent materials behavior, corrosion, fretting, wear, and other related phenomena of concern in all steam generators.The prototype Peach Bottom Unit No. 1 40-MW (electric) HTGR was operated by the Philadelphia Electric Company for a total of 1349 equivalent full power days during a 7-yr period. Upon planned decommissioning of that plant, the forced-recirculation U-tube steam generators and other components were subjected to destructive properties tests and metallurgical examinations. These tests and examinations showed the steam generators to be in very satisfactory condition.The 330-MW(electric) Fort St. Vrain HTGR, owned and operated by Public Service Company of Colorado, and now in the final stages of startup, has achieved 70% power and generated more than 1.5 × 106 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, and their design and development required considering a number of new materials factors including creep fatigue. Also, because of the once-through design, water chemistry control needed special consideration.Current designs of larger HTGRs also feature steam generators of helical tube once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of very large tubesheet forgings, consideration of the gaseous corrosion effects of the primary coolant, and other related factors.

A Three-Dimensional Depletion Analysis of the First Shippingport Pressurized Water Reactor

The first Shippingport seed-blanket core was operated for 5530 equivalent full power hours at equilibrium xenon and samarium conditions. The comparison of physics measurements and calculations presented are those applicable to the firs: core containing the initial seed material. A three- dimensional diffusion theory depletion analysis indicates that this calculational model describes with reasonable accuracy the directly observed and irferred reactor parameters examined over core lifetime. The reactor parameters compared include: criticality, reactivity lifetime, xenon transient behavior, temperature coefficients, and blanket power fraction. While the primary emphasis is on the three-dimensional calculational and experimental comparisons, the results of one and twodimensional diffusion theory depletion calculations are included to indicate their relative merit. The results indicate that such reactor parameters as excess reactivity, temperature coefficients, and blanket power fraction, may be estimated to within approximately the same accuracy by one and two-dimensional depletion models as by this par ticular three-dimensional model. This conclusion must be qualified by noting the crudeness employed in the threedimensional depletion model. (auth)