Void Worth Research Articles

Monte Carlo Modelling of Increasing Void Fraction in 100% MOX ABWR: Lessons Drawn from the FUBILA Program

The French Atomic Energy Commission CEA and the Japanese Incorporated Administrative Agency JNES (Japan Nuclear Energy Safety Organisation) have undertaken first-of-a-kind full MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache Centre. The experiments have been designed to obtain core physics data under high-burn-up 9 × 9 and 10 × 10 BWR MOX assemblies operating conditions. The experimental program, consisting of eight different core configurations, started in January 2005 and ended on September 1, 2006. The analysis of the void increase part of the experimental data between 0 and 70% void has been carried out using the French TRIPOLI-4.5 continuous-energy Monte Carlo calculation code with the newly released JEFF3.1.1 nuclear data library. The average C/E discrepancies obtained on critical masses, reactivity worth, and pin-by-pin power distributions enable us to estimate all the integral and local parameters with uncertainties largely within the target uncertainties, demonstrating the capability of the code to treat complex geometries with a high degree of accuracy. Additional keff calculations performed with the latest ENDF/B-VII evaluation exhibit a clear tendency to overestimate the keff by about 500 to 650 pcm and the void worth by more than 4%, showing that the JEFF3.1.1 library is more precise for MOX lattices.

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl–KCl–MgCl 2 (LSFR), sodium (SFR), and supercritical CO 2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO 2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO 2 reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.

Thermal hydraulic challenges of Gas Cooled Fast Reactors with passive safety features

Transient response of a Gas Cooled Fast Reactor (GFR) coupled to a recompression supercritical CO 2 (S-CO 2) power conversion system (PCS) in a direct cycle to a Loss of Coolant Accident (LOCA) and a Loss of Generator Load Accident is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post-LOCA events when system pressure is lost and when reliance on passive decay heat removal (DHR) is emphasized. Although it is possible to design a S-CO 2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal that provide flow in well defined regimes with low uncertainty, and can be easily overdesigned to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.

Annular Fast Reactor Cores with Low Sodium Void Worth for TRU Burning

Annular sodium-cooled fast reactor cores [600 MW(electric)] with low sodium void worth are developed for burning transuranic nuclides discharged from light water reactors. The several core design variants are developed by changing the core configuration, the core height, the fuel assembly design and type of nonfuel assemblies in the core, and their core performance parameters including safety-related reactivity coefficients are analyzed and inter-compared. The study focuses on the core neutronic parameters without going into the detailed safety and material compatibility studies. The study shows that the several cores of the annular type can be designed to have low sodium void worth, high transmutation capability, and all the negative temperature reactivity coefficients except for the positive one related to coolant expansion that can be compensated for by the reactivity coefficients by the fuel axial expansion and the fuel Doppler effects under the off-normal events, which increase temperatures. Of the cores considered, the use of a larger central control region and fuel assemblies with high coolant flow area in the core boundaries is found to be the most effective and simple way to achieve low sodium void worth and high transmutation capability.

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.

Studies of an Accelerator-Driven Transuranium Burner with Hafnium-Based Inert Matrix Fuel

Neutronic and burnup characteristics of an accelerator-driven transuranium burner in a startup mode were studied. Different inert and absorbing matrices as well as lattice configurations were assessed in order to identify suitable fuel and core design configurations. Monte Carlo transport and burnup codes were used in the analyses. The lattice pin pitch was varied to optimize the source efficiency and coolant void worth while respecting key thermal and material-related design constraints posed by fuel and cladding. A HfN matrix appeared to provide a good combination of neutronic, burnup, and safety characteristics: maintaining a hard neutron spectrum, yielding acceptable coolant void reactivity and source efficiency, and alleviating the burnup reactivity swing. A conceptual design of a (TRU,Hf)N fueled, lead-bismuth eutectic-cooled accelerator-driven system was developed. Twice higher neutron fission-to-absorption probabilities in Am isotopes were achieved compared to reactor designs relying on ZrN or YN inert matrix fuel. The production of higher actinides in the fuel cycle is hence limited, with a Cm fraction in the equilibrium fuel being ~40% lower than for cores with ZrN matrix-based fuel. The burnup reactivity swing and associated power peaking in the core are managed by an appropriate choice of cycle length (100 days) and by core enrichment zoning. A safety analysis shows that the system is protected from instant damage during unprotected beam overpower transient.

BWR核設計コードによるMOX臨界試験EPICUREおよびMISTRALの解析

For the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores.

Neutronics of Minor-Actinide Burning Accelerator-Driven Systems with Ceramic Fuel

We have investigated neutronic properties of lead-bismuth-cooled accelerator-driven systems with different minor-actinide-based ceramic fuels (two composite oxides and one solid-solution nitride). Adopting a transuranic composition with 40% plutonium in the initial load, transmutation rates of higher actinides (americium and curium) equal to 265 to 285 kg/GW(thermal)·yr are obtained. The smallest reactivity swing is provided by the magnesium oxide-based cercer fuel. The cercer cores, however, exhibit large coolant void worths, which is of concern in the case of gas bubble introduction into the core. Nitride and cermet cores are more stable with respect to void formation. The poorer neutron economy of the molybdenum-based cermet makes it difficult, however, to accommodate an inert matrix volume fraction exceeding 50%, a lower limit for fabricability. Higher plutonium fraction is thus required for the cermet, which would lead to lower actinide burning rates. The nitride core yields high actinide burning rates, low void worths, and acceptable reactivity losses.

05/00627 Coolant void worth in fast breeder reactors and accelerator-driven transuranium and minor-actinide burners

05/00627 Coolant void worth in fast breeder reactors and accelerator-driven transuranium and minor-actinide burners

Coolant void worth in fast breeder reactors and accelerator-driven transuranium and minor-actinide burners

Liquid metal coolant void worth have been calculated as a function of fuel composition and core geometry for several model fast breeder reactors and accelerator-driven systems (ADSs). The Monte Carlo transport code MCNP with continuous energy cross-section libraries was used for this study. With respect to the core void worth, lead/bismuth cooled FBRs appear to be inferior to those employing sodium for pitch-to-diameter ratios exceeding 1.4. It is shown that in reactor systems cooled by lead/bismuth eutectic radial steel pin reflector significantly lowers the void worth. The void worth proves to be a strong function of the fuel composition, reactor cores with high content of minor actinides in fuel exhibiting larger void reactivities than systems with plutonium based fuel. Enlarging the lattice pitch in ADS burners operating on Pu rich fuel decreases the void worth while the opposite fact is true for ADSs employing americium based fuels.

Neutronic aspects of inert matrix fuels for application in ADS

Accelerator driven systems may operate on uranium or thorium free fuels. In order to guarantee the stability of such fuels at high temperatures, the use of inert matrices is foreseen. In the present study, safety parameters of 800 MWth ADS cores operating on oxide and nitride fuels with high americium content are investigated for a representative range of pin and core geometries. It is shown that among the inert matrices investigated, chromium yields the lowest void worth, hafnium nitride the highest fission probability for americium and magnesia the highest burnup potential.

Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel

So-called dedicated fuels will be utilized to obtain maximum transmutation and incineration rates of minor actinides (MAs) in accelerator-driven systems (ADSs). These fuels are characterized by a high-MA content and the lack of the classical fertile materials such as 238U or 232Th. Dedicated fuels still have to be developed; however, programs are under way for their fabrication, irradiation, and testing. In Europe, mainly the oxide route is investigated and developed. A dedicated core will contain multiple “critical” fuel masses, resulting in a certain recriticality potential under core degradation conditions. The use of dedicated fuels may also lead to strong deterioration of the safety parameters of the reactor core, such as, e.g., the void worth, Doppler or the kinetics quantities, neutron generation time, and βeff. Critical reactors with this kind of fuel might encounter safety problems, especially under severe accident conditions. For ADSs, it is assumed that because of the subcriticality of the system, the poor safety features of such fuels could be coped with. Analyses reveal some safety problems for ADSs with dedicated fuels. Additional inherent and passive safety measures are proposed to achieve the required safety level. A safety strategy along the lines of a defense approach is presented where these measures can be integrated. The ultimate goal of these measures is to eliminate any mechanistic severe accident scenario and the potential for energetics.

Optimization of Height-to-Diameter Ratio for an Accelerator-Driven System

The height-to-diameter (H/D) ratio of a lead-bismuth eutectic (LBE)-cooled accelerator-driven system (ADS) has been evaluated in terms of the neutron multiplication, the coolant void worth, and the coolant velocity. For a model ADS, an optimization of the H/D ratio is performed with a Monte Carlo code both for the effective multiplication factor keff and for the multiplication of the external neutrons. In the optimization, ten cases of H/D values have been analyzed for a homogeneous fuel blanket. Also, the dependency of the optimal H/D ratio on the target/buffer is addressed. The Monte Carlo simulations show that the optimal H/D configuration of the ADS core is quite different for the two important measures, and a high H/D ratio can provide a significantly higher source multiplication than the traditional pancake core. Furthermore, various core analyses including depletion calculations are conducted for three selected heterogeneous cores with different H/D ratios, which are a small H/D value (pancake type), a medium H/D value, and a high H/D value, respectively. Void reactivity coefficients of the LBE coolant are evaluated and compared for the three designs to quantify the effects of the H/D ratio. Additionally, a thermal-hydraulic analysis has been performed to derive a maximum allowable core height subject to the LBE velocity limit due to its corrosion and erosion characteristics. It is shown that the practically optimal H/D ratio for source multiplication is tightly constrained by the maximum allowable LBE velocity, depending on the core design parameters.

Analysis of MISTRAL Experiments with JENDL-3.2

Nuclear Power Engineering Corporation has been analyzing measurement data of MISTRAL, a LWR MOX core physics experimental program, with a deterministic code system of the diffusion and the transport theories, SRAC, and a Monte Carlo transport code, MVP, based on JENDL-3.2. The program consists of one reference U02 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. The measurement parameters cover critical masses, boron concentrations, radial and axial core power distribution, spectrum indexes, conversion factors, effective delayed neutron fractions, iso-thermal temperature coefficients, boron efficiency, single absorber worth, water hole worth, absorber cluster worth and void worth. Critical keff calculations show a slight overestimate for MOX cores and a small systematic trend with aging of MOX fuel. Power distribution calculations show good agreement with the measurements in both homogeneous and mock-up cores. Measured spectrum indexes of fissile isotopes are well reproduced by the calculation. For other measurements, calculated results well reproduce the measurements within one or two times of the experimental error. The analysis verifies the validity of JENDL-3.2 and the analysis methods for the high moderation LWR MOX cores.

Application of Burnable Absorbers in an Accelerator-Driven System

The application of burnable absorbers (BAs) to minimize power peaking, reactivity loss, and capture-to-fission probabilities in an accelerator-driven waste transmutation system has been investigated. Boron-10-enriched B4C absorber rods were introduced into a lead-bismuth-cooled core fueled with transuranic (TRU) discharges from light water reactors to achieve the smallest possible power peakings at beginning-of-life (BOL) subcriticality level of 0.97. Detailed Monte Carlo simulations show that a radial power peaking equal to 1.2 at BOL is attainable using a four-zone differentiation in BA content. Using a newly written Monte Carlo burnup code, reactivity losses were calculated to be 640 pcm per percent TRU burnup for unrecycled TRU discharges. Comparing to corresponding values in BA-free cores, BA introduction diminishes reactivity losses in TRU-fueled subcritical cores by ~20%. Radial power peaking after 300 days of operation at 1200-MW thermal power was <1.75 at a subcriticality level of ~0.92, which appears to be acceptable, with respect to limitations in cladding and fuel temperatures. In addition, the use of BAs yields significantly higher fission-to-capture probabilities in even-neutron-number nuclides. Fission-to-absorption probability ratio for 241Am equal to 0.33 was achieved in the configuration studied. Hence, production of the strong alpha-emitter 242Cm is reduced, leading to smaller fuel-swelling rates and pin pressurization. Disadvantages following BA introduction, such as increase of void worth and decrease of Doppler feedback in conjunction with small values of βeff, need to be addressed by detailed studies of subcritical core dynamics.

A pan-shape transuranic burner core with a low sodium void worth

A 1575 MWt transuranic (TRU) burner reactor core with a low sodium void worth has been developed by devising a pan-shaped active core design. The core consists of two types of fule subassemblies that differ in the height of the fueled regions. This strategy has allowed an extreme "pancaking" of the inner core region while the radial dimension increase is limited by placing longer fuels in the outer core region. The fuel cycle analysis has been performed in the equilibrium cycle consisting of external feed fuel with reprocessed typical PWR spent fuel and fissile makeup with recycled TRU elements. The neutronic performance characteristics obtained from the equilibrium cycle analysis show that it would work safely as well as economically as measured in terms of burnup reactivity swing peak power density Doppler coefficient TRU burning and sodium void worth. The core has relatively low double power peaks in both the inner and outer cores without enrichment zoning and this enables to make an active core volume smaller. The developed TRU burner core has been subjected to an extensive parametric study on the reprocessing schemes. Investigations are given for sodium void worth transmutation capability burnup reactivity swing and minor actinide (MA) contents. Through this series of study a variant of the TRU burner core that is aimed at preferentially burning MA has been determined. This MA burner core uses U-235 as well as the homogeneously recycled TRU elements. This MA burner core consumes 140 kg of MA per year without penalizing the sodium void reactivity observed in the TRU burner core. Finally the combined introduction of developed TRU and MA burner cores shows the functional effectiveness of reducing PWR discharged TRU inventory.

A Liquid-Metal Reactor for Burning Minor Actinides of Spent Light Water Reactor Fuel—II: Nuclear Data Uncertainty Analysis

A comprehensive sensitivity and uncertainty analysis was performed on a 1200-MW(thermal) minor actinide burner designed for a low burnup reactivity swing, negative Doppler constant, and low sodium void worth. Sensitivities of the performance parameters were generated using depletion perturbation methods for the constrained closed fuel cycle of the reactor. The uncertainty analysis was performed using the sensitivity and covariance data taken from ENDF/B-V and other published sources. The uncertainty analysis of a liquid-metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. The relative uncertainties in the burnup swing, Doppler constant, and void worth were conservatively estimated to be 220, 120, and 59%, respectively. An analysis was performed to prioritize the minor actinide reactions for reducing the uncertainties.

A Liquid-Metal Reactor for Burning Minor Actinides of Spent Light Water Reactor Fuel—I: Neutronics Design Study

A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics.

The Central Void Reactivity in the Oak Ridge National Laboratory Enriched Uranium (93.2) Metal Sphere

The reactivity worth of a central void region in the Oak Ridge National Laboratory (ORNL) unmoderated and unreflected uranium (93.20 wt% 235U) metal sphere was obtained by replacement measurements in a small (0.460-cm-diam) central spherical region in this 3.4420-in.-radius sphere. The measured central void region worth was 9.165 ± 0.023 ¢ using the delayed neutron parameters of Keepin, Wimett, and Zeigler to obtain the reactivity from the measured stable reactor periods. This value is slightly larger than measurements for GODIVA I with larger cylindrical samples of uranium (93.70 wt% 235U) in the center: 135.50 ± 0.12 ¢/mol for GODIVA I and 138.05 ± 0.34 ¢/mol for the ORNL sphere measurements. The difference could be due to sample size effect. The central worth was also calculated by neutron transport theory methods to be 6.02 ± 0.01 × 10-4 Δk. The measured and calculated values are related by the effective delayed neutron fraction. The value of the effective delayed neutron fraction obtained in this way from the ORNL sphere is 0.00657 ± 0.00002, which is in excellent agreement with that obtained from GODIVA I measurements, where the effective delayed neutron fraction was determined as the increment between delayed and prompt criticality and was 0.0066. From these ORNL measurements, using the delayed neutron parameters of ENDF-B/VI to obtain the reactivity from the stable reactor period measurements, the central void worth is 7.984 ± 0.021 ¢, and the inferred effective delayed neutron fraction is 0.00754. These values are 14.2% higher than those obtained from use of the Keepin, Wimett, and Zeigler delayed neutron data and produce a value of effective delayed neutron fraction in disagreement with GODIVA I measurements, thus questioning the usefulness of the six-group delayed neutron parameters (fast fission) of uranium from ENDF-B/VI for obtaining the reactivity from the measured reactor period using the Inhour equation.