Both Boiling Water Reactor Research Articles

A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies

The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (∼hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. We present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.

Characterization of oxides formed on iron-chromium-aluminum alloy in simulated light water reactor environments

AbstractTo avoid accidents like that in Fukushima, the US Department of Energy is engaged with a nuclear fuel vendor to evaluate the performance of iron-chromium-aluminum (FeCrAl) alloys such as advanced powder metallurgy tubing (APMT) as accident-tolerant material for uranium dioxide fuel cladding in light water reactors (LWR). It was important to characterize the oxides formed on APMT under both boiling water reactor (BWR) and pressurized water reactor (PWR) environments for a better understanding of its environmental sustainability in LWRs. Coupons of APMT were exposed for 1 year to both hydrogenated and oxygenated high-purity water at 288°C (e.g. simulated BWR water chemistry without Pt injection) and hydrogenated high-purity water at 330°C (e.g. simulated primary PWR water chemistry without Li/B addition). Results show that after 1-year immersion, APMT always developed a chromium-rich protective oxide film on its surface. In oxygen-containing environments, the oxide consisted of a dual layer, an external thicker layer containing mostly iron oxides and a thinner internal layer rich in chromium oxide. In hydrogen environments, only a single oxide layer formed, consisting of chromium oxide. This is a similar finding as for type 304 and 316 stainless steels and for nickel-based alloy 600, which is extensively reported in the literature. General corrosion of APMT alloys under LWR operating conditions would not be a limiting factor for its performance as cladding material.

Evaluation of Helical-Cruciform Fuel Rod Assemblies for High-Power-Density LWRs

Because of the immense capital costs associated with new nuclear construction, interest remains high in developing strategies to uprate existing light water reactors (LWRs) for higher power density and in raising core power density for next-generation LWR designs. Toward these goals, the helical-cruciform (HC) fuel rod assembly has been proposed. The HC fuel rod assembly is a self-supporting nuclear fuel configuration consisting of four-petaled, axially twisted fuel rods closely packed in a square array. Advantages over traditional fuel geometry include a larger surface-to-volume ratio and improved radial mixing characteristics. The self-supporting nature of the assembly obviates the need for grid plates, improving core hydraulics.Past studies have identified these and other benefits of HC fuel rod geometry and have adapted its shape and design to LWR fuel assemblies for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. However, because of a lack of suitable thermal-hydraulic correlations to capture HC rod bundle flow behavior, this work fell short of a complete assessment of the potential. Recent progress has been made in this regard due to the empirical development of specialized hydraulic and lateral mixing correlations for HC rod geometry. As a result, advanced LWR core designs taking advantage of the HC fuel rod assembly can be reexamined with a greater degree of precision and confidence.For a BWR core using HC rod assemblies, applying the new HC rod bundle correlations to subchannel models uncovered a need to increase the hydraulic diameter of the tight side and corner subchannels, to prevent flow starvation. Small protrusions were added to the assembly box side at axial locations corresponding to each rod quarter-twist to act as spacers. This prompted a slight redesign of the rod cross-sectional shape. Likewise, the central water rod region was adjusted to maintain the reference hydrogen-to-uranium atom ratio. With these changes, subchannel models predicted a 24% allowable power uprate for the 200-cm twist pitch HC core, in comparison to a reference BWR with traditional fuel. The uprate is accomplished assuming a fixed-core power-to-flow ratio.In comparison, modeling showed that a PWR core employing HC fuel rod assemblies may allow power uprates up to 47%, for a fixed power-to-flow ratio. One major difference from the BWR case is that subcooled critical heat flux (CHF) levels rise with increasing coolant mass velocity, opposite the trend for saturated CHF conditions. However, subcooled CHF is also known to be more sensitive to locally peaked heat flux, which was not explicitly modeled in these simulations. Power density gains claimed here will be ultimately dependent on the degree to which the HC rod’s twist disrupts nascent pockets of vapor as subcooled CHF limits are approached; this effect should be further investigated experimentally.

Characterization of 14C in Swedish Light Water Reactors

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

パッケージリアクタの開発

A new small reactor concept named Package-Reactor has been developed through a joint research of Hitachi, Ltd., and Mitsubishi Heavy Industries. Several key design of its nuclear steam supply system have been investigated, taking into account both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) technologies. The PackageReactor is a stand-alone energy supply system, and is designed to attain high reliability, high safety, good maintainability, good operability and low construction cost. To achieve these aims, the reactor adopts natural-circulation core cooling systems. The reactor has no active devices inside its high pressure boundary. Combining a turbine electric power generation and biomass refining, which is supported by JGC Corporation or chemical heat pipe systems attains a perfect base load operation. The whole system is simple and small to be easily constructed with a very short period even at remote regions with poor infrastructures. The Package-Reactor is an innovative nuclear power plant concept to pioneer and develop new markets of the nuclear power business.

Development of new transportable storage casks for interim dry storage

AbstractWe have started a programme to design a new type of transportable storage cask (Hitz casks) for both boiling water reactor (BWR) and pressurised water reactor (PWR) fuels for use in the new interim dry spent fuel storage project in Japan. The basic policy of this development is to use proven technology to realize a safe and cost-effective design with a high transport and storage capacity and a low fabrication cost. Since it is not permissible to change the lid gaskets at the storage facility, the double-lid system is designed to be able to use double metallic gaskets as the containment boundary for transport after the storage period; this is one of the new design features used in the casks. With the basket design we tried to achieve a capacity of 69 fuel assemblies for BWR fuel and 26 fuel assemblies for PWR fuel. Further details about these and other topics are discussed.

Optimization of MOX enrichment distributions in typical LWR assemblies using a simplex method-based algorithm

The enrichment distributions within light water reactor (LWR) fuel assemblies are optimized using a modified linear programming (SIMPLEX Method) technique initiated from a flat enrichment distribution until a target, maximum local power peaking factor is achieved. The optimum rod enrichment distribution when each rod is allowed to have its own individual enrichment (for this case the target, maximum local power peaking factor is 1.0) is obtained at an intermediate point of the optimization procedure. Later, the optimal locations and values for a reduced number of rod enrichments (groups) are obtained for an input target maximum local power peaking factor by applying sensitivity to change techniques. After an initial set of enrichment groups have been defined, interchanges of rods among neighboring groups are carried out to obtain the final assembly enrichment distribution. The optimization procedure is demonstrated by presenting results for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel assembly designs. Reactor-grade plutonium (with a fissile Pu fraction of 59.6%) and weapons-grade plutonium (with a fissile Pu fraction of 94.0%) were assumed to be the feed Pu material for the MOX fuel rods in the BWR and PWR examples, respectively. Hot-full-power-temperature and beginning-of-life conditions were also assumed in the example problems.

Experimental investigation of flow transient critical heat flux at light water reactor conditions

An experimental investigation of flow transient critical heat flux (CHF) for water at pressures 6.9 and 15.5 MPa is conducted under the simulation of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) conditions. The influence of flow decay rates on local mass velocity and quality at CHF is explored to ensure the adequacy of local-condition approach in the CHF predictions. The assesment of transient upstream correction factor to modify the predictions of steady-state CHF correlations during the BWR flow transients are presented. A mechanistic CHF model based on the dryout of a sublayer and CHF correlations are evaluated in the predictions of PWR flow transient CHF.

Behavior of Breached Light Water Reactor Spent Fuel Rods in Air and Inert Atmospheres at 229°C

A whole-rod test was conducted at 229 °C to investigate the long-term stability of light water reactor spent fuel rods with artificially induced defects stored in inert and unlimited air atmospheres. Both boiling water reactor (BWR) and pressurized water reactor (PWR) rods were tested. After 2235 h, visual observations, diametral measurements, and radiographic smears were used to assess the degree of cladding deformation and particulate release. The same examinations plus metallography and x-ray analysis were conducted after 5962 h. Neither of the breached rods tested in inert atmosphere, nor the breached PWR rod tested in unlimited air, showed any measurable change from the pretest condition. The upper defect on the BWR rod tested in unlimited air had a 12.7-mm split after 2235 h and had 10% cladding deformation. The crack grew to 63.5 mm after 5962 h. X-ray analysis indicated that the UO2 had oxidized to U3O8 at this defect. The difference in the behavior of the upper and lower defects is attributed to the accessibility of the air to the fuel due to the positioning of the defect with relation to the pellet/pellet interface.

Design and Operational Characteristics of High-Density Fuel Storage Facilities

General Electric (GE) experience in operation of the Morris spent fuel storage facility, which now contains over 300 Mg of both boiling water reactor (BWR) and pressurized water reactor spent fuel, confirms that receipt, handling, and storage of spent fuel can be accomplished safely with negligible impact on the environment or the operation itself. Basin water treatment is accomplished with disposable powdered resins applied to a precoated filter-demineralizer unit, and special applications of Zeolites aid in maintaining radiocobalt and radiocesium concentrations to <4 × 10−4 μCi/ml in the basin water. No gaseous radioisotopes from damaged or leaking fuel have been observed, and no significant increases in radioactivity or loss of cladding integrity have been observed during fuel handling and storage.GE has utilized this experience to design an expansion of the Morris basin and to design Boral-poisoned, high-density, stainless-steel storage modules for BWR reactor pools. These free-standing modules store BWR fuel on 165.1-mm (6.5-in.) center spacing, and a sliding low-friction support system limits the seismic loads applied to the fuel. Application of this fuel storage experience has permitted expansion of storage capacity for spent fuel at Morris and at BWR reactors, permitting continued operation until federal programs for long-term storage have been clarified and implemented.