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At least 973 records · Page 54

Thermal Evaluations for Hypothetically Drain-Down Spent Fuel Storage Facility at SRS

The L-Area Spent Nuclear Fuel (SNF) underwater storage facility at Savannah River Site (SRS), referred to as L-Basin SNF facility, has an array of Vertical Tube Storage (VTS) racks and a group of High Flux Irradiation Reactor (HFIR) fuel cores. The underwater facility is integrally designed as a wet SNF storage system with a chemically-controlled water pool. The SNF storage racks are seismically qualified. The L-Basin pool holds about 3.4 million gallons of water with concrete wall thickness of about 5 ft and water depths of 17 ft to 30 ft. The facility has mainly two different elevation levels of 30 ft and 17 ft below the ground level. At the level 30 ft below the ground, a maximum of 110 racks of VTS and 5 racks of Dry Cave can be stored. At the level of 17 ft below the ground, 120 HFIR cores can be stored, considering that each rack contains two HFIR cores. In case of a hypothetical accident, all cooling water in the storage pool may be drained away, resulting in dry SNF storage condition. The primary objective of the work is to estimate the maximum fuel temperature in a conservative manner when the L-Basin facility is completely dry due to the hypothetical drain-down of the pool water. A three-dimensional Computational Fluid Dynamics (CFD) modeling approach was taken for the thermal evaluations of the water-drained SNF storage facility. The modeling calculations were performed, assuming that all of the storage racks in L-Basin are fully loaded with SNF assemblies containing decay heat sources. In the case of full loading, the L-Basin facility stores 14,600 SNF assemblies in VTS racks and 120 cores in HFIR racks. The thermal analysis results of the L-Basin model yielded maximum temperatures of VTS and HFIR fuel assemblies for different fuel loading scenarios and boundary conditions of the SNF storage racks.

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Improving Experience Beyond the Ride: Joint Assignment and Repositioning for Rideshare

In this study, a novel joint assignment and repositioning strategy is designed, implemented, and tested in a simulated traffic environment. The strategy addresses limitations of previously constructed rideshare repositioning algorithms by integrating the consideration of rider pooling into the optimization. By considering pooling, the strategy can minimize excess vehicle movement and increase riders' use of pooled services over a zonal-flow-based repositioning strategy and a fleet operating with no repositioning at all. The new strategy can increase pooling by up to 2.3% compared to no repositioning in a large region, and up to 0.1% in a small region, without compromising overall fleet revenue. Additionally, compared to a baseline, flow-based repositioning strategy, the new strategy can reduce empty vehicle travel by up to 30% in the larger region, and 2.1% in the smaller simulated area.

Paul, Joseph↗

In-situ Observations of Directed Energy Deposition Additive Manufacturing Using High-Speed X-ray Imaging

In laser-based directed energy deposition (DED) additive manufacturing, interactions among the laser beam, particle flow, and melt pool influence the properties of the solidified final part. Two separate DED systems, one with high powder flow rates to represent industrial-scale DED processing and the other with low powder flow rates for individual particle tracking, were synchronized with the high-speed imaging setup at the Advanced Photon Source in Argonne National Laboratory. In-situ x-ray imaging of the DED process using both systems highlighted the influence of powder flow rates. Increased powder flow rates resulted in less laser absorption into the melt pool, leading to a transition from a keyhole mode to a melt pool without a keyhole but with surface fluctuations due to powder flow. Increased velocities of particles during powder flow resulted in a decrease in particle melting times and a greater propensity for porosity formation. Altogether, better understanding of the interactions that occur during various scales of the DED process will enable flexibility, control, and new materials development in DED-based additive manufacturing.

36 MATERIALS SCIENCE↗

Role of scan strategies on thermal gradient and solidification rate in electron beam powder bed fusion

Local microstructure control in electron beam powder bed fusion (EB-PBF) is of great interest to the additive manufacturing community to realize complex part geometry with targeted performance. The local microstructure control relies on having a detailed understanding of local melt pool physics (e.g., 3-D melt pool shape as well as spatial and temporal variations of thermal gradient (G) and solidification rate (R)). In this research, a new scan strategy referred to as ghost beam is numerically evaluated as a candidate to achieve the targeted G and R of IN718 alloy. The boundary conditions for simulations, including the speed (490 mm/s) and spatial locations of the beam within a given layer, are obtained by using series of snapshot images, recorded at 12,000 frames per second, using a high-speed camera. The heat transfer simulations were performed using TRUCHAS an open-source software deployed within a high-performance computational infrastructure. The simulation results showed that reheating at short beam on-time and time delay decreases both G and R. Local variation of R at the center of the melt pool trailing edge showed periodic temporal fluctuations. Finally, the ghost beam scan strategy was compared to other existing raster and spot scan strategies.

36 MATERIALS SCIENCE↗

Predictive process mapping for laser powder bed fusion: A review of existing analytical solutions

One of the main challenges in the laser powder bed fusion (LPBF) process is making dense and defect-free components. These porosity defects are dependent upon the melt pool geometry and the processing conditions. Power-velocity (PV) processing maps can aid in visualizing the effects of LPBF processing variables and mapping different defect regimes such as lack-of-fusion, under-melting, balling, and keyholing. This work presents an assessment of existing analytical equations and models that provide an estimate of the melt pool geometry as a function of material properties. The melt pool equations are then combined with defect criteria to provide a quick approximation of the PV processing maps for a variety of materials. Finally, the predictions of these processing maps are compared with experimental data from the literature. Here, the predictive processing maps can be computed quickly and can be coupled with dimensionless numbers and high-throughput (HT) experiments for validation. The present work provides a boundary framework for designing the optimal processing parameters for new metals and alloys based on existing analytical solutions.

36 MATERIALS SCIENCE↗

Performance estimation of resonance-enhanced dual-buoy wave energy converter using coupled time-domain simulation

This article presents the modeling methodology and performance evaluation of the resonance-enhanced dual-buoy WEC (Wave Energy Converter) by HEM (hydrodynamic & electro-magnetic) fully-coupled-dynamics time-domain-simulation program. The numerical results are systematically compared with the authors’ 1/6-scale experiment. With a direct-drive linear generator, the WEC consists of dual floating cylinders and a moon-pool between the cylinders, which can utilize three resonance phenomena from moon-pool dynamics as well as heave motions of inner and outer buoys. The contact and friction between the two buoys observed in the experiment are also properly modeled in the time-domain simulation by the Coulomb-friction model. Moon-pool resonance peaks significantly exaggerated in linear potential theory are empirically adjusted through comparisons with measured values. A systematic comparative study between the simulations and experiments with and without PTO (power-take-off) is conducted, and the relative heave displacements/velocities and power outputs are well matched. Then, parametric studies are carried out with the simulation program to determine optimum generator parameters. The performance with various wave conditions is also assessed.

16 TIDAL AND WAVE POWER↗

Application of underwater spectrometric system for survey of ponds of the MR reactor (NRC Kurchatov institute)

The underwater spectrometric system for survey the bottom of material science multi-loop reactor MR ponds was elaborated. This system uses CdZnTe (CZT) detectors that allow for spectrometric measurements in high radiation fields. The underwater system was used in the spectrometric survey of the bottom of the MR reactor pool, as well as in the survey located in the MR storage pool of highly radioactive containers and parts of the reactor construction. As a result of these works irradiated nuclear fuel was detected on the bottom of pools, and obtained estimates of the effective surface activity detected radionuclides and created by them the dose rate. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

A life cycle and product type based estimator for quantifying the carbon stored in wood products

Background: Timber harvesting and industrial wood processing laterally transfer the carbon stored in forest sectors to wood products creating a wood products carbon pool. The carbon stored in wood products is allocated to end-use wood products (e.g., paper, furniture), landfill, and charcoal. Wood products can store substantial amounts of carbon and contribute to the mitigation of greenhouse effects. Therefore, accurate accounts for the size of wood products carbon pools for different regions are essential to estimating the land-atmosphere carbon exchange by using the bottom-up approach of carbon stock change. Results: To quantify the carbon stored in wood products, we developed a state-of-the-art estimator (Wood Products Carbon Storage Estimator, WPsCS Estimator) that includes the wood products disposal, recycling, and waste wood decomposition processes. The wood products carbon pool in this estimator has three subpools: (1) end-use wood products, (2) landfill, and (3) charcoal carbon. In addition, it has a user-friendly interface, which can be used to easily parameterize and calibrate an estimation. To evaluate its performance, we applied this estimator to account for the carbon stored in wood products made from the timber harvested in Maine, USA, and the carbon storage of wood products consumed in the United States. Conclusion: The WPsCS Estimator can efficiently and easily quantify the carbon stored in harvested wood products for a given region over a specific period, which was demonstrated with two illustrative examples. In addition, WPsCS Estimator has a user-friendly interface, and all parameters can be easily modified.

54 ENVIRONMENTAL SCIENCES↗

Vibro-acoustic Imaging at the Breazeale Reactor

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spent TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Deep Convection: Onset, Diurnal Cycle, and Role in Climate Variability and Sensitivity (Final Report)

The ASR SGP testbed for continuous modeling offers new opportunities to improve GCM parameterizations. We conducted an ASR investigation of convective and cloud microphysical processes that used existing ARM data as a testing ground for meaningful approaches to model evaluation and set the stage for later strategic use of the testbed to improve cumulus parameterizations and better understand deep convective cloud feedback. Our research consisted of the following tasks: Diurnal cycle of continental convection Deep convection at the SGP occurs in two forms: Organized mesoscale convection that originates upstream and produces peak precipitation at night; and local shallow convection that develops into deep convection that peaks in late afternoon. Local transitions are a logical next step from the shallow convection that will be the early focus of the testbed. This transition constrains entrainment and cold pools, since GCMs produce a continental rain peak near noon. We studied ARSCL-observed transition cases using the WRF, determined the sources of SCM diurnal cycle errors and made parameterization changes as appropriate. Factors controlling convective detrainment and anvil evolution Convective detrainment is a major source of high cloud ice water content and cloud forcing. The GISS GCM diagnoses a convective updraft speed profile, assumes a particle size distribution (PSD), and uses size-fallspeed relations to determine the fraction of condensate carried upward vs. precipitating. We used SCM case studies for TWP-ICE and MC3E for which vertical velocity has been retrieved to test whether entrainment improvements produce realistic updraft speeds and then improved the updraft microphysics parameterization. We compared detrained ice water contents and fall speeds to ARM and other field experiment data to refine the GCM’s fall speed and PSD assumptions. Role of convective cloud radiative processes in the Madden-Julian Oscillation (MJO) Changes in entrainment and rain evaporation and a cold pool parameterization have produced an MJO in the GISS GCM for the first time and allowed us to reproduce the KAZR dependence of convection depth on humidity during AMIE-Gan. We determined whether recent model changes including cold pools improve or degrade the MJO and made adjustments accordingly.

54 ENVIRONMENTAL SCIENCES↗

Fire tests and analyses of a rail cask-sized calorimeter.

Three large open pool fire experiments involving a calorimeter the size of a spent fuel rail cask were conducted at Sandia National Laboratories Lurance Canyon Burn Site. These experiments were performed to study the heat transfer between a very large fire and a large cask-like object. In all of the tests, the calorimeter was located at the center of a 7.93-meter diameter fuel pan, elevated 1 meter above the fuel pool. The relative pool size and positioning of the calorimeter conformed to the required positioning of a package undergoing certification fire testing. Approximately 2000 gallons of JP-8 aviation fuel were used in each test. The first two tests had relatively light winds and lasted 40 minutes, while the third had stronger winds and consumed the fuel in 25 minutes. Wind speed and direction, calorimeter temperature, fire envelop temperature, vertical gas plume speed, and radiant heat flux near the calorimeter were measured at several locations in all tests. Fuel regression rate data was also acquired. The experimental setup and certain fire characteristics that were observed during the test are described in this paper. Results from three-dimensional fire simulations performed with the Cask Analysis Fire Environment (CAFE) fire code are also presented. Comparisons of the thermal response of the calorimeter as measured in each test to the results obtained from the CAFE simulations are presented and discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Corrosion test of LWR cladding tubes in high temperature seawater

In the case of cooling system failure in spent fuel pools (SFP) due to station black out and/or loss of coolant accidents, a pool water level must be kept high enough to avoid overheat of spent fuels. Seawater is one of the candidate coolant materials in such emergencies. In the present work, we have studied the corrosion behavior of zirconium alloys in seawater. Tube samples of Zircaloy-2 and -4 claddings were immersed in artificial seawater (3, 10 and 20 wt.%) at high temperature (80 - 85 Celsius degrees). Boric acid was added to the seawater at a corrosion test of Zircaloy-4. Two types of specimens were prepared according to the simulation of circumferential tensile stress. One was a plugged tube specimen, which was pressurized to around fuel rod pressure in a spent fuel pool by argon gas. Another specimen was of C-ring shape, where large circumferential tensile stress was partially induced at the surface. A notch was longitudinally grooved to increase the local tensile stress on some of the C-ring specimens. The specimen surface was observed by a scanning electron microscopy (SEM) after the corrosion test. After the 200 day long corrosion test for the plugged tube specimens without boric acid, Mg(OH){sub 2} deposits were observed on the surface of Zircaloy-2 and -4 specimens whereas Mg(OH){sub 2} was not observed with boric acid because of probably a lower pH. Neither pitting corrosion nor stress corrosion cracking (SCC) was observed on the surface of all plugged tube specimens. Only in the C-ring specimens in 20 wt.% seawater without boric acid and in 3 and 10 wt.% seawater with boric acid, small cracks were observed at the notch. The results of the present work show that Zircaloy-2 and -4 have a high tolerance towards corrosion and SCC in high temperature seawater.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

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System impacts of accelerated transfer of used nuclear fuel to dry storage - 15220

In the wake of the September 11, 2001 terror attacks, and again after the March 11, 2011, accident at the Fukushima Dai-Ichi station from the Great East Japan Earthquake/tsunami, concern was expressed regarding security of current spent fuel storage practices in the United States. An approach recurrently proposed to reduce the potential for fire and the release of radionuclides in an accident or sabotage scenario is to reduce the loading of the reactor fuel pools, which are frequently maintained near their full capacity at many reactor sites in the U.S. We assess the impact on the waste management system of different scenarios for accelerating the schedule of used nuclear fuel (UNF) transfer from reactor pools to dry storage casks. We consider scenarios where the spent fuel pools are drawn down to 50% or 25% of full capacity over 5- or 10-year schedules. The scenarios include status quo, where UNF continues to remain in dry storage at the reactor sites, as well as scenarios with UNF transfer to interim storage and eventually to a mined geologic repository. The primary purpose is to evaluate how the different scenarios for accelerating the transfer of UNF to at-reactor dry storage would affect the performance of the entire UNF management system. Areas that were considered include an estimate of the additional system costs that could be incurred and where the existing infrastructure at the reactor sites would come under stress from the transfer demands. This study considered only the fuel-handling logistics and rough-order-of-magnitude cost at reactor sites, interim storage facility, and operations to package UNF into disposal canisters. The effects of minimum fuel age that could be transferred and the accelerated transfer starting date are also examined, but found to have relatively little effect. Impacts on safety and worker dose are not evaluated quantitatively, although effects can be inferred from increases in handling operations, etc. from the scenarios considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Robotic Underwater Decontamination Services for Nuclear Applications-18464

Successful Underwater Robotic Decontamination Demonstrates Dose-Saving And Outage-Shortening Technology Diakont successfully provided underwater robotic decontamination services on a North American nuclear power plant's refueling cavity and dryer-separator pool in Spring 2017 using a new tool system. Historically, nuclear plant operators have conducted cleaning and decontamination of these surfaces manually, after draining water from the space. However manual decontamination is slow, and can result in excessive personnel dose exposure. The underwater robotic decontamination services present a vast improvement over manual decontamination because it reduces personnel dose exposure, reduces radwaste, doesn't impact plant chemistry, and doesn't risk inadvertently spreading contamination. Also the new decontamination method avoids the risk of personnel injury and component damage associated with hydrolasing. And by performing the decontamination robotically while the cavities are flooded, in many cases the critical path outage schedule can be shortened. Able to decontaminate horizontal, vertical, and curved surfaces, the decontamination tools easily navigate to areas within flooded cavities that are inaccessible to previous solutions, performing decontamination in parallel to other activities including fuel movement. Also, unlike other legacy solutions, the new tools do not require continuous use of an overhead crane or other method of suspension while performing the decontamination activities. The ROV-type decontamination tool attaches and drives along the cavity and component surfaces using a high-force, no-flow vortex generator, even in the presence of Residual Heat Removal (RHR) or shutdown cooling flow. Efficient, effective cleaning is performed using a rugged brushing action to detach the crud, while vacuuming it away at high flow rates to a submerged filter. At the Spring 2017 refueling outage, the decontamination services with the new methodology were so effective that no additional manual cavity decontamination was required after drain-down. Using the tool's ability to swim, attach, and crawl, field technicians decontaminated the majority of the surfaces designated by the plant operator. Preliminary surveys indicated that all contamination levels were reduced to <50 K dpm/100 cm{sup 2}, helping the utility meet their INPO/Industry collective radiation exposure goals. The new cleaning and decontamination services are ideal for servicing various underwater areas within nuclear power plants such as refueling cavities, BWR-6 drywell heads, spent fuel pools, cask loading pools and fuel transfer canals. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of k{sub eff} in the unfolding process of the LOTUS reactor design

An innovative small thermal reactor concept named LOTUS, which is a pool-type reactor aimed at heat supply. The passive concept applied in the LOTUS and it has a separable core. It means that its core can passively unfold into four parts along its central axis only with the buoyancy and gravity when some accidents happen, especially the wall rupture of the pool caused by earthquake, aircraft impact and so on. This idea can effectively enhance the ability of pool-type reactor for resisting some severe accidents. We have executed some pre-computations in order to check a non-supercritical system for the transient simulation. In this paper, the development of k{sub eff} have been calculated by both stepwise steady-state and transient modes using Serpent, as well as including some sensitivity analyses about the k{sub eff} to unfolding velocity, inserting velocity of control rod (CR) and time step. The result showed that k{sub eff} value decreased definitely as the reactor unfolded and the consistent change exposed between the stepwise steady state and transient result when the unfolding angle varied from 0 to 1 degree. However, the stepwise steady state k{sub eff} negatively declined after that angle, which demonstrated that for such a dynamic unfolding change the transient mode was more suitable, peculiarly for considering delayed neutrons and neutron moderation. For sensitivity calculation, the k{sub eff} decreased more rapidly as the unfolding velocity and inserting velocity of CR increased, and a smaller time step corresponded to a more stable change. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗