Light Water Breeder Reactor Research Articles

Neutronics calculation of the conceptual TRISO duplex fuel rod design

• The neutronic properties of TRISO duplex fuel rod for a small block-type thorium HTR was investigated. • TRISO duplex fuel rod consists of an inner region filled with UO 2 particles and an outer layer filled with ThO 2 particles. • TRISO duplex fuel increases the discharged burnup and has a lower reactivity swing. • TRISO duplex fuel capable of maintaining a softer neutron spectrum throughout its burnup period. The duplex fuel pellet includes UO 2 components as seed and ThO 2 blanket, originally designed to increase the breeding rate of fissile materials in light water breeder reactors. However, the seed portion alone concentrates most of the energy released from the fission reaction, which can make the seed temperature too high. The duplex fuel potential for the high-temperature reactor (HTR) using TRISO fuel has never been researched. MCNPX simulation was conducted to compare the neutronics performance of normal TRISO fuel with that of TRISO duplex fuel. Two types of HTR fuel rod, the cylinder and the annular rod, were simulated and compared to see whether the duplex neutron characteristics would be better in the cylinder or annular fuel compact design. The TRISO duplex rods produced higher burnup than normal rods up to 5% higher for cylinder and 6% for annular rods. The TRISO duplex rod is also capable of maintaining longer criticality, producing smaller k inf changes and reducing the potential for plutonium build-up. The proposed TRISO duplex conceptual fuel design has an interesting neutronic properties, which offers possibilities for future studies.

Validation of the Monte Carlo code MONK for thermal 232Th/233U systems using the JEF-2.2 and JEFF-3.1 libraries

There is limited validation of MONK using 232Th/233U benchmarks, and existing validation of Monte Carlo codes has generally used older nuclear data libraries. Here we use MONK-9A to systematically study a series of thermal 232Th/233U-based benchmark experiments, taken from the International Criticality Safety Benchmark Evaluation Project handbook and using JEF-2.2 and JEFF-3.1 libraries. We find that 232Th-dominated thermal systems (those moderated with graphite or light water) with limited 233U content show little difference when modelled with JEF-2.2 or JEFF-3.1. However, when there is significant 233U then JEFF-3.1 is found to give considerably better agreement with experiment. There is likely to be little benefit in utilising JEFF-3.1 instead of JEF-2.2 when modelling systems for which the predominant nuclides are 232Th, 235U and 238U. However, for systems containing large fractions of 233U – such as light water breeder reactors operating on a 232Th-233U cycle – it is recommended that JEFF-3.1 is utilised.

The TRANSURANUS burn-up model for thorium fuels under LWR conditions

The Transuranus Burn-up (TUBRNP) model for (Th,Pu) O2 and (Th,U) O2 fuels is described and validated. To this end, the depletion equations of the nuclides 232Th, 233U and 234U were included as well as the breeding of 235U through a neutron capture in 234U. The Monte Carlo code Serpent was utilized to derive the one-group effective cross sections and the radial power-shape functions that account for the resonance absorption in the epithermal range of neutron energies in 232Th and 240Pu. Finally, TUBRNP is completed by extending the fission yields of 233U for the most representative isotopes of Nd, Xe, Kr and Cs. The validation of TUBRNP for (Th,Pu) O2 type fuels was carried out in two steps. A comparison of the normalised radial distributed concentrations of Th, U, Pu, Nd, Xe and Cs between TUBRNP, Serpent and Electron probe microanalysis (EPMA) data points measured on a sample from a rodlet irradiated at Kernkraftwerk Obrigheim (KWO) allows checking the correctness in the derivation of the radial functions. In a second step, a comparison of the radially averaged values of the same elements between TUBRNP, Serpent, EPMA and benchmark codes under same conditions indicates the agreement in the computation of the one-group cross sections. For (Th,U) O2 fuels isotopic compositions measured in a rod segment from the Light Water Breeder Reactor (LWBR) Shippingport irradiation programme were employed to validate the calculation of the 1-group cross sections in TUBRNP for this type of fuel.

Criticality Benchmarks for the Epithermal Test Assembly Experiments

The Epithermal Test Assembly (ETA) experiments were performed to test the adequacy of 233U, 235U, and 232Th cross sections in epithermal spectra in support of the Light Water Breeder Reactor (LWBR) Program. The ETA design contained a central heavy water–moderated test region surrounded by a light water–moderated annular driver region. Two series of experiments were performed: ETA-I with 235UO2-ThO2 fuel rods in the test region and ETA-II with 233UO2-ThO2 fuel rods in the test region. The dominant uncertainties in the critical configurations include the test-rod pitch, fuel density, and ThO2 mass for ETA-I; the test-region fuel-rod fuel density and 233U to (233U + Th) weight ratio for ETA-II; and the driver-region fuel-rod outer diameter, uranium enrichment, and pitch for both ETA experiments. Benchmark model results using MCNP5 are provided for ENDF/B-V, ENDF/B-VI, ENDF/B-VII.0, and ENDF/B-VII.1 cross sections with only the ENDF/B-VII.0 results falling within three standard deviations of the benchmark model keff. The ETA-I and ETA-II benchmark evaluations have been included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and are replicated in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Viability of uranium nitride fueled high-conversion PWR

Light water breeder reactors have been examined periodically for decades, and considerable research into the neutronics and thermal hydraulics of several designs has been performed. Currently, there is an interest in reduced moderation boiling water reactors in Japan and in the United States, where Department of Energy (DOE)-sponsored university research programs are examining proposed breeding BWRs.This study assesses the thermal hydraulic and neutronic feasibility of a nitride fueled pressurized water reactor (PWR) breeder design. Because of the higher fuel density, the use of nitride fuel would be preferable to the traditional oxide fuel for a high conversion PWR design. To illustrate the advantage, a design that uses large hexagonal assemblies was considered. The design has 14 inner seed pin rows and 4 outer blanket pin rows, leading to 505 inner seed pins and 366 outer blanket pins per assembly. Each assembly contains 6 thimble locations for control rods. The same assembly type can be used at all locations in the core. For this study a fuel pin diameter of 1.2 cm was used with a gap of 1 mm between pins. The seed regions had 12.75% percent of the heavy metal as fissile plutonium (Pu-239 and Pu-241), hosted in depleted uranium (0.25% U-235). The plutonium vector for the seed region was 2.7% Pu-238, 47.9% Pu-239, 30.3% Pu-240, 9.6% Pu-241, 8.5% Pu-242 and 1.0% Am-241. The Blanket regions contained only depleted uranium with 0.25% U-235.This assembly was modeled with the Monte Carlo depletion code Serpent to determine the initial to final fissile inventory ratio (FIR). The as-specified assembly model did not achieve an FIR value above 1.0, so the water in the thimble regions was changed to void, to lower the H/HM ratio. This led to FIR values above 1.0 for the oxide, the 85% theoretical density nitride (N85) and the 95% theoretical density nitride (N95). All designs had an FIR of 1.03 at 35 MWd/kgHM. However, the single batch discharge burnup of the assembly was much higher for the nitride cases. The burnups were 15.6, 24.5 and 32.2 MWd/kgHM for the oxide, N85 and N95 respectively.The KfK correlation for critical heat flux was applied to calculate the minimum departure from nucleate boiling ratio (MDNBR), whose value was set to be 1.45. With an acceptable core pressure drop of 400 kPa and acceptable vibration magnitude, the maximum power in an assembly and the maximum flow rate were determined. A nitride fuel had a larger MDNBR than the oxide fuel. This increase is likely a result of spectrum hardening, which was found to reduce the fractional amount of power generated in the seed region, where departure from nucleate boiling (DNB) occurs. Because of the longer neutron mean free path in the harder spectrum, more fission events occur in the blanket. The coolant density coefficient of the nitride fuel was found to have positive coefficient values in voided thimble tube cases. However, the power coefficient was negative and has the ability to limit power increases. The radially reflective assembly boundary may have made the analysis more conservative than full core analysis.

The Behavior of ThO2-Based Fuel Rods during Normal Operation and Transient Events in LWRs

The thermal, mechanical, and chemical behavior of both thorium and uranium dioxide (ThO2-UO2) and thorium and plutonium dioxide (ThO2-PuO2)-based fuels during in-service and hypothetical accident conditions in light water reactors (LWRs) is described. These fuels offer the possibility for increased proliferation resistance and a reduction in the stockpile of weapons-grade and reactor-grade PuO2 as well as being a more stable waste form. The behavior is described for three different designs of ThO2-based fuels: a homogeneous mixture of ThO2-UO2, a microheterogeneous arrangement of the ThO2 and UO2, and a homogeneous mixture of ThO2-PuO2. The behavior was calculated with widely known LWR analysis tools extended for ThO2-based fuels: (a) MATPRO for calculating material properties, (b) FRAPCON-3 for calculating in-service fuel temperature and fission-gas release, (c) VIPRE-01 for calculating the possibility for departure from nucleate boiling, (d) HEATING7 for calculating in-service two-dimensional temperature distributions in microheterogeneous fuel, (e) SCDAP/RELAP5-3D for calculating the transient reactor system behavior and fuel behavior during loss-of-coolant accidents, and (f) FRAP-T6 for calculating the vulnerability of the cladding to cracking due to swelling of the fuel during hypothetical reactivity-initiated accidents.The analytical tools accounted for the following differences in ThO2-based fuels relative to 100% UO2 fuel: (a) higher thermal conductivity, lower density and volumetric heat capacity, less thermal expansion, and higher melting point; (b) higher fission-gas production for 233U fission than 235U fission, but a lower gas diffusion coefficient in the ThO2 than in the UO2; (c) less plutonium accumulation at the rim of the fuel pellets; (d) greater decay heat; (e) microheterogeneous arrangement of fuel; and (f) more-negative moderator temperature and Doppler coefficients and a smaller delayed-neutron fraction. The newly developed models for ThO2 were checked against data from the light water breeder reactor program. Calculations by these analytical tools indicate that the in-service and transient performance of homogeneous ThO2-UO2-based fuels with respect to safety is generally equal to or better than that of 100% UO2 fuel. The in-service and transient temperatures in the most promising neutronic design of microheterogeneous ThO2–UO2–based fuel are greater than the temperatures in 100% UO2 fuel but are still within normal LWR safety limits. The reactor kinetics parameters for ThO2-PuO2–based fuel cause a higher transient reactor power for some postulated accidents, but in general, the margin of safety for ThO2-PuO2 fuels is equal to or greater than that in 100% UO2 fuels.

A Fission Gas Release Model for High-Burnup LWR ThO2-UO2 Fuel

Fission gas release in thoria-urania fuel has been investigated by creating a specially modified FRAPCON-3 code. Because of the reduced buildup of 239Pu and a flatter distribution of 233U, the new model THUPS (Thoria-Urania Power Shape) was developed to calculate the radial power distribution, including the effects of both plutonium and 233U. Additionally, a new porosity model for the rim region was introduced at high burnup. The mechanisms of fission gas release in ThO2-UO2 fuel are expected to be essentially similar to those of UO2 fuel; therefore, the general formulations of the existing fission gas release models in FRAPCON-3 were retained. However, the gas diffusion coefficient was adjusted to a lower level to account for the smaller observed release fraction in the thoria-based fuel. To model the accelerated fission gas release at high burnup properly, a new athermal fission gas release model was introduced. The modified version of FRAPCON-3 was calibrated using the measured fission gas release data from the light water breeder reactor. Using the new model to calculate the gas release in typical pressurized water reactor hot pins gives data that indicate that the ThO2-UO2 fuel will have considerably lower fission gas release above a burnup of 50 MWd/kg HM.

Dissolution Behavior and Fission Product Release from Irradiated Thoria-Urania Fuel in Groundwater at 90°C

ABSTRACTThe dissolution behavior and fission-product release from irradiated thoria-urania fuel was studied by immersing fuel samples in J-13 well water at 90°C. The samples are from the Shippingport Light Water Breeder Reactor and consist of binary solid solutions of (U,Th)O2 with UO2 contents varying from 2.0 to 5.2 Wt.%. The post-irradiation U isotopic composition of the samples used in our experiments is: 87.3% 233U, 10.4% 234U, 1.8% 235U, and <0.5% 238U, 236U, 232U. Burn up values for the samples range from 22.3 to 40.9 megawatt-days per kg-metal. Our tests were performed on polished disks and on crushed and sieved samples in stainless-steel reaction vessels with air-filled head-space. After 196 days of reaction, samples showed no evidence for corrosion at the micrometer scale. Concentration ranges (μgL-1) of key radionuclides in filtered (∼5 nm pore size) leachates were: 0.1 – 15 90Sr, 0.9 – 7.0 99Tc, 0.1 – 35.2 137Cs,<0.2 – 0.8 233U, <0.1 – 0.7 232Th. Concentrations of 237Np, 239Pu, 240Pu and 241Am were all <0.2 μgL-1. The relatively high concentrations of the fission products 90Sr and 137Cs occur early during leaching and decrease for later samplings. Matrix dissolution rates for the irradiated thoria-urania samples range from ∼3x10-3 to <3×10-5 mg m-2day-1 and are at least two orders of magnitude lower than those measured for UO2 spent fuels under similar experimental conditions.

Physics Experiments and Lifetime Performance of the Light Water Breeder Reactor

The light water breeder reactor (LWBR) operated at the Shippingport Atomic Power Station from 1977 to 1982, serving the electric power grid for the Greater Pittsburgh area. The LWBR was a pressurized water reactor (PWR) with several unique features: It was designed and proved to be a breeder with an end-of-life fissile fuel content ∼1.3% greater than beginning of life; the reactor used the 233U-Th fuel system; and it had a large Doppler coefficient, low reactivity worth of transient xenon, and a significant reactivity effect from transient 233Pa. There were no control rods or soluble poison, and reactivity was controlled by movable fuel.Core operations went extremely well. The design lifetime of 18 000 effective full-power hours was exceeded by 60% by utilizing a gradual reduction in power level. The overall capacity factor was 65%.Physics experiments showed good agreement with predictions of movable fuel reactivity worth, most temperature coefficients, breeding, power distribution, and xenon stability. Unexpected results occurred in measurements of flow coefficient of reactivity, zero power temperature coefficients early in life, and bred fissile fuel distribution.The LWBR technology has demonstrated that water-cooled breeder reactors can operate in existing water power plants much like conventional PWRs.

A Gauge for Nondestructive Assay of Irradiated Fuel Rods

The light water breeder reactor (LWBR) was developed by the Bettis Atomic Power Laboratory under the technical direction of the Office of Naval Reactors, U.S. Department of Energy. The LWBR operated successfully in the Shippingport Atomic Power Station from 1977 to 1982, producing more than 29,000 equivalent full-power hours. Because of the small breeding gain (1.35%) predicted for this self-sustaining breeder, proof of breeding required accurate nondestructive assay of expended fuel from the LWBR. The fact that breeding has been proven in the LWBR means that this reactor provides a vast alternative energy resource using plentiful thorium fuel. A gauge, called the production irradiated fuel assay gauge (PIFAG), was developed to nondestructively assay whole irradiated fuel rods from the LWBR. The gauge uses the method of active interrogation with /sup 252/Cf source neutrons and delayed neutron counting. A description is given of the PIFAG, its calibration, and its application to the assay of irradiated LWBR fuel rods. A comparison of the PIFAG results with destructive assay results for 17 irradiated LWBR fuel rods show that the two methods are in excellent agreement, differing by 0.069 and 0.162% for the low- and high-energy PIFAG interrogation spectra, respectively.

The High Gain Light Water Breeder Reactor with a Uranium-Plutonium Cycle

In the design concept presented, two seed-blanket cores are utilized, a prebreeder and a breeder. The prebreeder core is fueled with plutonium obtained from standard LWR spent fuel and generates plutonium with a high isotopic content of /sup 240/Pu and /sup 241/Pu, which is used to fuel the breeder core. The initial fissile fuel for a 950-MW(electric) prebreeder is between 3000 and 4000 kg. Assuming the availability of rapid fuel reprocessing and refabrication, overall breeding of well over 10%/6 yr is feasible. The time between core refueling is >1 yr, rather than once every 3 months as required in the case of liquid-metal fast breeder reactors, so that the fuel inventories are substantially reduced for our system. Thermal-hydraulic analysis indicates that each of the two seed-blanket cores can fit into a standard pressurized water reactor pressure vessel and meet safety requirements. Major advantages of the concept are continued utilization of present LWR plants and strong negative void coefficients for the core. It is anticipated that at equilibrium conditions, when plutonium for the prebreeder and breeder cores is recycled into standard LWRs, there will be a further gain in breeding.

Utilization of Thorium in Light Water Reactors

An alternative method of thorium utilization in light water reactors (LWRs) is proposed. The main idea of the proposed concept is to apply a different fuel management scheme for the neutron-producing part of the core, the uranium seed, and for the neutron-absorbing part of the core, the thorium blanket. An example of the specific design based on this concept was analyzed, and preliminary evaluation indicated the potential of significant savings in uranium consumption. The fuel cycle of the proposed concept includes reprocessing and refabrication of uranium fuel only, without separation of plutonium and /sup 233/U isotopes. Such a fuel cycle offers higher proliferation resistance compared with the LWR recycle mode of operation or the light water breeder reactor fuel cycle. Finally, the feasibility of the reactor design based on the proposed concept may be established after detailed thermal-hydraulic analysis and study of the irradiation behavior of the thorium-based fuel.

The Reactivity Worth of Protactinium-233 Inferred from Measurements

In a thorium-fueled reactor, the conversion of /sup 232/Th to /sup 233/U involves the intermediate nuclide /sup 233/Pa. This isotope has a significant reactivity effect in any thorium reactor, especially one with an important epithermal flux component. The light water breeder reactor, operating in the Shippingport atomic power station, is a /sup 233/U-Th reactor with about half the power produced at energies above thermal. The reactivity effect of full-power equilibrium /sup 233/Pa has been inferred from critical position measurements and control element reactivity worths to be approximately 2.5% delta rho in this reactor, confirming calculational predictions.

A German Reaction to U.S. Nonproliferation Policy

A s an aid in understanding the views of a citizen of the Federal Republic of Germany (FRG) on U.S. nonproliferation policy, it is useful at the outset to review the energy situation in the FRG, which in many respects is representative of the situation in other European countries. Nuclear energy plays an important part in this country's endeavors to reduce dependence on oil imports, to diversify sources of supply, and to generate electrical power as economically as possible. Two constraints apply: (1) because the FRG has no significant uranium resources at its disposal but is fully dependent upon imports from non-European countries, it must use reactor systems and fuel cycles permitting maximum utilization of fuel; and (2) the disposal of power plant waste, including spent fuel assemblies, must be effected in a permanent and environmentally acceptable manner. The Government, public opinion, and the courts are making their approval of further development of nuclear energy conditional on a satisfactory long-term solution to the waste management problem. This solution, which also satisfies the first of the above constraints, consists of closing the uranium-plutonium fuel cycle by reprocessing the spent fuel assemblies and manufacturing new mixed oxide fuel assemblies for use in light water or fast breeder reactors; that is, the deliberate development of a plutonium economy.1 The German nuclear fuel cycle center, which is in the project planning stage, will satisfy both of the constraints mentioned above and will also do much to serve the aims of nonproliferation. As all systems are concentrated at one site, the transportation of nuclear fuels is minimized; since the plutonium made available by reprocessing is immediately made up into new fuel assemblies, the amount of free, explosive-grade fissile material is kept as small as possible; by recycling into reactors, the plutonium is conveyed to safe and well-protected locations where it is transformed into non-fissile material by nuclear fission; and international surveillance of the nuclear fuel center will ensure timely discovery of any plutonium diversion. Nuclear energy is also one of the Federal Republic's crucial exports. Even in the fossil-fueled sector, the German electrical industry is highly dependent upon exports in order to ensure the full utilization of its engineering and manufacturing

Authors

Authors

Measurement and Analysis of Parameters in Tight 232Th-235U and 232Th-233U Lattices Moderated with Heavy Water

Parameter measurements and calculations in two D2O-moderated thorium-uranium critical assemblies are described in detail. The first, designated ETA-I, contained 6.7 wt% 235UO2-ThO2 fuel rods 0.66 cm in diameter, clad in aluminum. The second assembly, ETA-II, contained 3.0 wt% 233UO2-ThO2 fuel rods 1.09 cm in diameter, clad in Zircaloy-2. A relatively hard spectrum was obtained in both lattices. Because leakage was small and the central flux spectra were closely asymptotic, these assemblies provided a cleanly analyzable test of important reaction cross sections. Parameters measured in ETA-I were 232Th fission/235U fission (δ02), 232Th capture/235U fission (CR*), and epithermal/thermal-neutron ratios for 235U fission (δ25) and 232Th capture (p02). Corresponding measurements with 233U were made in ETA-II; δ02 was measured directly in fuel and ThO2 pellets by observing specific fission product gamma rays with a high resolution Ge(Li) detector. With this system, 232Th capture gamma-ray activity was also measured directly in a fuel pellet. The epithermal/thermal-neutron ratios were measured by the thermal-neutron subtraction technique relative to 164Dy, and CR* was measured relative to its thermal-neutron value in a very soft spectrum. A three-dimensional adjoint Monte Carlo program was used to calculate foil flux perturbations and cadmium cutoff energies. Supplemental fast/epithermal-neutron spectrum comparisons were made between the ETA assemblies and the 4/1 TRX critical assembly by means of the following activation detectors: 197Au(n,γ), 55Mn(n,γ), 235U(n,f), 238U(n,f), 27Al(n,α), and 96Zr(n, γ). The lattices were analyzed explicitly with a full energy range Monte Carlo program having a detailed cross-section description. ENDF/B Version 2 and light-water breeder reactor (LWBR) cross-section sets were used. On the whole, agreement with experiment was very good.

Use of Metallic Thorium for LWBRs and LWRs

The potential of metallic thorium as a reactor fuel has been examined. Two particular cases were taken into consideration: a light water breeder reactor (LWR) and a pressurized water reactor (PWR). The comparison with a standard LWBR (ThO/sub 2/) and with a standard PWR (UO/sub 2/) points out the potential benefits of the proposed concepts from the economic point of view, as well as from that of conserving natural resources. (13 references) (auth)