bluuebunny/crossref_metadata_2025 huggingface dataset:
Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX
Authors: Ho¨hne Thomas , Kliem So¨ren , Rohde Ulrich , Weiß Frank-Peter
Published: September 2008

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.
Financial Benefits of Process Data Reconciliation in Power Generating Plants
Authors: Jansky Andy
Published: September 2008

Process Data Reconciliation (PDR) is a certified method that calculates the most likely values considering process measurement uncertainties and closing all energy- and material balances where all interdependencies within the entire plant process are fulfilled in a covariance matrix. There are three main factors that generate the financial benefits for the user of reconciled data, depending on the type of plant and base/peak load behaviour: • Increased efficiency / maximized output; • Time advantage in retrieving “lost” megawatts; • Reduction of maintenance costs.
Design and Testing of D.C. Conduction Pump for Sodium Cooled Fast Reactor
Authors: Nashine B. K. , Dash S. K. , Gurumurthy K. , Rajan M. , Vaidyanathan G.
Published: September 2008

DC Conduction pump immersed in sodium forms a part of Failed Fuel Location Module (FFLM) of 500 MWe Fast Breeder Reactor (PFBR) currently under construction. FFLM housed in control plug of the reactor, is used to locate the failed fuel sub-assembly due to clad rupture in the fuel pin. The DC conduction pump sucks the sodium from the top of fuel sub-assemblies through the selector valve and pumps the sodium to hold up for detecting the presence of delayed neutrons. Presence of delayed neutron is the indication of failure in the sampled fuel sub-assembly. The DC Conduction Pump was chosen because of its low voltage operation (2 V) where argon/alumina ceramic can provide required electrical insulation even at operating temperature of 560°C without much complication on the manufacturing front. Sampling of sodium from top of different sub-assemblies is achieved by operation of selector valve in-conjunction with the drive motor. FFLM requires the pump to be immersed in sodium pool at ∼560°C located above the fuel sub-assemblies in the reactor. The Pump of 0.36 m3/h capacity and developing 1.45 Kg/ cm2 pressure was designed, manufactured and tested. The DC Conduction Pump has a stainless steel duct filled with liquid sodium, which is to be pumped. The stainless steel duct is kept in magnetic field obtained by means of electromagnet. The electromagnet is made of soft iron and the coil made of copper conductor surrounds the yoke portion of electromagnet. The external DC source of 2000 Amps, 2 Volt is used to send current through sodium placed in the stainless steel duct and the same current is sent through copper coil of electromagnet for producing required magneto motive force, which in turn produces required magnetic field. The interaction of current in sodium (placed in stainless steel duct) and magnetic field produced by the electromagnet in the duct region produces pumping force in the sodium. Electromagnet, copper coil, stainless steel duct, copper bus bar etc. are encapsulated in stainless steel shell. Hydraulic characteristics, efficiency, cavitation free operation at operating temperatures was ascertained by conducting tests in sodium loop called Large Component Test Rig (LCTR). The pump was also endurance tested for 750 hrs. The performance tests on DC Conduction Pump indicate that the pump meets the target specification at reactor operating condition. This paper deals with design, construction and performance testing of DC Conduction Pump.
Integration of Passive System Reliability in PSA Studies
Authors: Burgazzi Luciano , Marques Michel
Published: September 2008

The treatment of passive safety systems within the probabilistic safety assessment models is a difficult and challenging task. The main concern arises from the nature of the passive systems whose predominant operating principles are based on physical phenomena rather than on active components. The present study provides a consistent approach for the integration of passive safety systems into fault tree and event tree based Probabilistic Safety Assessment (PSA) model of accident sequences, in the fashion of and in combination with a front line system or a human action. With reference to the thermal-hydraulic passive systems (e.g. natural circulation systems), in addition to the component failures (i.e. mechanical and electrical faults), the impairment of the physical principle upon which the system relies, deserves special consideration. This makes the relative assessment process different as regards the system model commonly adopted in the fault tree approach (i.e. exponential failure model). For the thermal-hydraulic passive system, since the failure process is driven mainly by the occurrence of the phenomenological failure modes, each pertinent basic event will be characterized by defined critical parameters (e.g. non-condensable fraction) that are expected to drive the failure mechanisms. An application of this approach is presented, with reference to a system designed for decay heat removal of advanced Light Water Reactors, relying on natural circulation and provided with a heat exchanger immersed in a cooling pool, acting as heat sink, and connected to the pressure vessel via steam and condensate lines.
Study for Nuclide Transfer Ratio of Particles Generated by Thermal Cutting
Authors: Iguchi Yukihiro , Baba Tsutomu , Kawakami Hiroto
Published: September 2008

In dismantling of nuclear facilities, secondary products would disperse into air and water. Thus, it is necessary to assess the effective dose for the public due to the gaseous and liquid radioactive wastes discharged. Especially for the cutting of a highly activated reactor core structures, the material may release special nuclides, some of which can be metal with low melting temperature and relatively high vapor pressure. In this case, the vaporized metal easily transfers to the particles instead of leaving in the dross and would make significant impact to the work place and environment. A pressure-tube type reactor often utilizes zirconium alloy as core structure, which may contains nuclides with high vapor pressure such as radioactive tin or antimony caused by activated tin. In this study, the radioactivity transfer ratio for thermal cutting in air has been investigated by cold tests. The tests have been carried out in an air-tight clean house and used a dust sampling system. The test piece was zirconium and niobium 2.5% alloy with some impurities, which was non-radioactive spare material for the pressure tube of Fugen NPS. Plasma cutting method was carried out with 200 A electric current. The particles were separated by one micrometer size with membrane filters and the material was investigated with an Energy Dispersion X-ray Analyzer (EDX) and the shape of the particle was observed with a Scanning Electron Microscope (SEM). The results showed that there was almost no tin in the dross because most of the tin was evaporated and transferred to aerosol. However, the ratio of zirconium, niobium and iron in the aerosol was not as high as that of tin. The vaporization pressure of antimony is higher than that of tin. The tin can be easily activated by neutron and becomes radioactive antimony. Because Zircalloy-2 or 4 contains ca. 1 percent of tin, irradiated Zircalloy contains significant radioactive antimony (Sb-125). Preparatory analyses showed that the Sb-125 to the environmental impact was comparable with other nuclides. For this reason a test for taking data of the transfer ratio of antimony by using irradiated material has been also carried out.
Overview of Seismic Fragility Capacity Test of PWR and BWR Control Rod Insertion Systems
Authors: Inagaki M. , Suzuki K. , Abe H.
Published: September 2008

In 2002FY the program of seismic fragility capacity test of PWR and BWR control rod insertion systems was initiated with the following objectives: • To examine the functional fragility of PWR and BWR control rod insertion systems using actual size mockup specimens for strong earthquake ground motions exceeding the design levels induced by the large shaking table. • To develop methodology of fragility evaluation and to establish fragility database of PWR and BWR control rod insertion systems. This paper is a report on an overview of the test program.
Neutronic and Physical Characteristics of an Accelerator Driven System with a Lead-208 Coolant
Authors: Khorasanov Georgy L. , Ivonov Anatoly P. , Blokhin Anatoly I.
Published: September 2008

In the paper a possibility of using a lead isotope, pure Pb-208, as a coolant for a subcritical core of 80 MW thermal capacity of the PDS-XADS type facility is considered. Calculations of neutronic characteristics were performed using Monte Carlo technique. The following initial data were chosen: an annular core with a target, as a neutron source, at its centre; the core coolant — Pb-208 (100%); a fuel — a mix of mono nitrides of depleted uranium and power plutonium with a small share of neptunium and americium; the target coolant — a modified lead and bismuth eutectic, Pb-208(80%)-Bi(20%); proton beam energy — 600 MeV; effective multiplication factor of the core under operation — Keff = 0.97; thermal capacity of the core — N = 80 MW. From calculations performed it follows that in using Pb-208 as the core coolant the necessary intensity of the external source of neutrons to deliver 80 MW thermal capacity is equal to S = 2.29−1017 n/s that corresponds to proton beam current Ip = 2.8 mA and beam capacity Pp = 1.68 MW. In using natural lead instead of Pb-208 as the core coolant, effective multiplication factor of the core in normal operating regime falls down to the value equal to Keff = 0.95. In these conditions multiplication of external neutrons in the core and thermal capacity of the subcritical core are below nominal by 1.55 times. For achievement the rated core power N = 80 MW it is required on ∼20–30% to increase the fuel loading and volume of the core, or by 1.55 times to increase intensity of the external source of neutrons. In the last case, the required parameters of the neutron source and of the corresponding proton beam are following: intensity of the neutron source S = 3.55·1017 n/s., beam current Ip = 4.32 mA, beam capacity Pp = 2.59 MW. To exploit the accelerator with the reduced proton beam current it will be required about 56 tons of Pb-208, as a minimum, for the core coolant. Charges for its obtaining can be recovered at the expense of the economy of the proton accelerator construction cost. In this case, the acceptable price of the lead isotope Pb-208 must be less than $2,860/kg.
A Techno-Economic Optimization of the Power Conversion System of a Very High Temperature Reactor
Authors: Mansilla Christine , Dumas Michel , Werkoff Franc¸ois
Published: September 2008

Generation IV nuclear reactors will not be implemented unless they enable lower production costs than with the current systems. In such a context a techno-economic optimization method was developed and then applied to the power conversion system of a very high temperature reactor. Techno-economic optimization consists in minimizing an objective function that depends on technical variables and economic ones. The advantage of the techno-economic optimization is that it can take into account both investment costs and operating costs. A techno-economic model was implemented in a specific optimization software named Vizir, which is based on genetic algorithms. The calculation of the thermodynamic cycle is performed by a software named Tugaz. The results are the values of the decision variables that lead to a minimum cost, according to the model. The total production cost is evaluated. The influence of the various variables and constraints is also pointed out.
Advances in Computational Fluid Dynamics Modeling of Two Phase Flow in a Boiling Water Reactor Fuel Assembly
Authors: Tentner Adrian , Lo Simon , Ioilev Andrey , Melnikov Vladimir , Samigulin Maskhud , Ustinenko Vasily , Kozlov Valentin
Published: September 2008

A new code, CFD-BWR [1], is being developed for the simulation of two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. CFD-BWR is a specialized module built on the foundation of the commercial CFD code STAR-CD [2] which provides general two-phase flow modeling capabilities. New models describing the inter-phase mass, momentum, and energy transfer phenomena specific for BWRs have been developed and implemented in the CFD-BWR module. A set of experiments focused on two-phase flow and phase-change phenomena has been identified for the validation of the CFD-BWR code and results of two experiment analyses focused on the radial void distribution are presented. The close agreement between the computed results, the measured data and the correlation results provides confidence in the accuracy of the models.
Design and Qualification of Control and Safety Rod and Its Drive Mechanism of Fast Breeder Reactor
Authors: Rajan Babu V. , Veerasamy R. , Rangaswamy D. , Narayanan K. , Pavan Kumar S. C. S. , Dash S. K. , Meikandamurthy C. , Rajan K. K. , Rajan M. , Puthiyavinayagam P. , Chellapandi P. , Vaidyanathan G. , Chetal S. C.
Published: September 2008

Prototype Fast Breeder Reactor (PFBR) has two shutdown systems. The absorber rod of the first system is called Control & Safety Rod (CSR). Control & Safety Rod Drive Mechanism (CSRDM) facilitates start-up & controlled shut-down of reactor and control of reactor power by raising and lowering of CSR and shutdown of the reactor on abnormal conditions by rapid insertion of CSR into the core, i.e., by scram action. After the detailed design and analysis of CSR and CSRDM, they were qualified in two stages. In the first stage, the critical assemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals, were tested individually simulating the operating conditions of the reactor and the design parameters were fine-tuned. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts. In the second stage, prototype CSRDM and CSR were manufactured and subjected to various functional tests in air, in hot argon and subsequently in sodium simulating the operating conditions of the reactor. Tests were carried out keeping CSRDM and CSR at aligned condition and with the possible misalignment between them. The performance was checked and recorded maintaining the temperature of sodium starting from 473 K to 823 K. Then the system was subjected to endurance tests. The results show that the performance of CSRDM and CSR is satisfactory and there is no significant change in the performance during endurance testing.
Primary Study of Steady States of HTGR-GT
Authors: Cao Jianhua , Wang Jie , Yang Xiaoyong , Yu Suyuan
Published: September 2008

The High Temperature Gas-cooled Reactor coupled with gas turbine (HTGR-GT) is supposed to be one of the candidates for the future nuclear power plants in both electricity and hydrogen production. The HTGR-GT cycle is theoretically based on the closed Brayton cycle with recuperator, intercooler and precooler. In this paper, the exergy analysis on a typical HTGR coupled with Gas Turbine was presented. Besides the core outlet temperature of the cycle, other operating parameters have been calculated to see their effect of the exergy efficiency of the whole cycle. The results show that, the compressor pressure ratio (γ) of the cycle has a great effect on the exergy losses of both heat-exchange and work components. And only the exergy loss of the recuperator decreases with the increase of γ, especially sharply when γ is small, while the exergy losses of other components increases. So there are optimized γ under different working conditions. The effect of other operating parameters, like pressure drop, recuperation efficiency and isentropic efficiencies of compressors and turbine, has been evaluated. The effect of two formers is similar, both obvious at the range of near the optimized γ, and γ goes down slightly when these two operating conditions are better. Compressors and turbine, as the work components of the cycle, their isentropic efficiency is quite important to the exergy efficiency of the cycle. The bigger γ, the more effect. Whatever, the isentropic efficiency of compressors and turbine is already quite high.
The DBA Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core
Authors: Lang Mingang , Dong Yujie
Published: September 2008

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, INET has planned to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basic accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the accident on one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was lager than 800°C, the reactor was scrammed. During the accident sequence the maximum fuel temperature was 1200.9°C. It was lower than the fuel temperature limitation of 1230°C. The paper compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10 when the fuel safety limits are the same.
Simulation of Hydrogen Distribution in Ignalina NPP ALS Compartments During BDBA
Authors: Babilas Egidijus , Urbonavicius Egidijus , Rimkevicius Sigitas
Published: September 2008

Accident Localisation System (ALS) of Ignalina NPP is a “pressure suppression” type confinement, which protects the population, employees and environment from the radiation hazards. According to the Safety Analysis Report for Ignalina NPP ∼110 m3 of hydrogen is released to ALS compartments during the Maximum Design Basis Accident. However in case of beyond design basis accident, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. If the volume concentration of hydrogen in the compartment reaches 4%, there is a possibility for a combustible mixture to appear. To prevent the possible hydrogen accumulation in the ALS of the Ignalina NPP during an accident the H2 control system is installed. The results of the performed analysis derived the places of the possible H2 accumulation in the ALS compartments during the transient processes and assessed the mixture combustibility in these places for a beyond design basis accident scenario. Such analysis of H2 distribution in the ALS of Ignalina NPP in case of BDBA was not performed before.
The BEMUSE Programme: Results of the First Part Concerning the LOFT L2-5 Test
Authors: de Cre`cy Agne`s , Bazin Pascal , D’Auria Francesco , Petruzzi Alessandro , Ryu Yong-Ho
Published: September 2008

This paper is aimed at describing results of the first part of the BEMUSE (Best Estimate Methods – Uncertainty and Sensitivity Evaluation) programme. The purpose of BEMUSE is the comparison of best-estimate calculations, followed by the comparison of uncertainty and sensitivity analyses for a Large Break Loss of Coolant Accident (LB-LOCA). The first part of the programme is devoted to the study of the LOFT L2-5 experiment. After a general presentation of the programme, which implies more than ten participants, this paper describes the qualification process and the results of the best-estimate calculations. The results are significantly less dispersed than those of the ISP-13, concerning already LOFT L2-5 more than 20 years ago. Then, it presents extensively the methods and the results of uncertainty and sensitivity analyses. All the participants, apart from the University of Pisa with the CIAU method, use a fully probabilistic approach, based on Wilks’ formula. However, differences appear for the choice of the uncertain input parameters to be considered and for their associated range of variation. Sensitivity analysis is performed with regression techniques, and the results are also compared. As a conclusion, main lessons learnt from BEMUSE and recommendations are presented.
Experimental Investigation of Evaporation Behavior of Polonium and Rare-Earth Elements in Lead-Bismuth Eutectic Pool
Authors: Ohno Shuji , Miyahara Shinya , Kurata Yuji , Katsura Ryoei , Yoshida Shigeru
Published: September 2008

Equilibrium evaporation behavior was experimentally investigated for polonium (210Po) in liquid lead-bismuth eutectic (LBE) and for rare-earth elements gadolinium (Gd) and europium (Eu) in LBE to understand and clarify the transfer behavior of toxic impurities from LBE coolant to a gas phase. The experiments utilized the ‘transpiration method’ in which saturated vapor in an isothermal evaporation pot was transported by inert carrier gas and collected outside of the pot. While the previous paper ICONE12-49111 has already reported the evaporation behavior of LBE and of tellurium in LBE, this paper summarizes the outlines and the results of experiments for important impurity materials 210Po and rare-earth elements which are accumulated in liquid LBE as activation products and spallation products. In the experiments for rare-earth elements, non-radioactive isotope was used. The LBE pool is about 330–670 g in weight and has a surface area of 4cm × 14cm. 210Po experiments were carried out with a smaller test apparatus and radioactive 210Po produced through neutron irradiation of LBE in the Japan Materials Testing Reactor (JMTR). We obtained fundamental and instructive evaporation data such as vapor concentration, partial vapor pressure of 210Po in the gas phase, and gas-liquid equilibrium partition coefficients of the impurities in LBE under the temperature condition between 450 and 750°C. The 210Po test revealed that Po had characteristics to be retained in LBE but was still more volatile than LBE solvent. A part of Eu tests implied high volatility of rare-earth elements comparable to that of Po. This tendency is possibly related to the local enrichment of the solute near the pool surface and needs to be investigated more. These results are useful and indispensable for the evaluation of radioactive materials transfer to the gas phase in LBE-cooled nuclear systems.
Solution to a Standard Thermal Hydraulics Problem in a Liquid Metal Subassembly
Authors: Son Hyoung M. , Suh Kune Y.
Published: September 2008

The model subassembly of the BREST-type reactor core consists of a pin bundle of square arrangement. In this bundle there are two zones which differ in the pin diameter and heat production. The model pin bundle contains one spacer grid which is located near the mid-plane of the rod bundle geometry. The working is a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Three kinds of experiments were performed to observe the thermal and hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperature distribution, central measuring pin simulator external surface temperature distribution, and coolant velocity distribution over the perimeter of the measuring pin simulator. The experiments were performed five times with increasing pin power ratios. Analysis was performed on the model subassembly of the BREST-type reactor core using a subchannel analysis code MATRA and a computational fluid dynamics code CFX. Calculational results were compared against the experimental data. The experiment revealed that the temperature rise was strongly dependent upon the geometry of the pin simulator. In contrast to the experimental results, the MATRA results were mainly dependent upon the thermal and hydraulic conditions. It was concluded that MATRA requires modifications for the pressure drop correlations that were considered inappropriate for accurately simulating the coolant behavior near the BREST-type grid spacer. Hand calculations were additionally carried out under different assumptions to determine the coolant exit temperature distribution in the pin simulator. First, the hand calculation was performed to find the coolant exit temperature distribution assuming that there is no momentum or energy transfer between subchannels. Second, an assumption was made that the coolant mixing in the subchannel assembly took place instantaneously and the pressure was equilibrated at the channel exit. Since MATRA is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. Here, CFX based on the finite volume method was utilized to calculate the velocity fields over the perimeter. Results from the experiment and CFX were averaged in each subchannel region so as to check on the tendency. The CFX analysis showed reasonable results which can be improved by imposing more detailed geometry accounting for the angle of the inclination of the grid spacer.
Preliminary Analysis of Phebus FPT3 Experiment with the Severe Accident ICARE2 Code
Authors: de Luze Olivier , Repetto Georges , Seiler Nathalie , Dominguez Christina
Published: September 2008

The last test of the international Phebus Fission Product (FP) program, named FPT-3, has been performed in November 2004 at Cadarache. It used boron carbide as absorber material, instead of Ag-In-Cd in the previous tests. Such absorber material is used in power plants like BWR, VVER, some western type PWR and EPR, and the importance of its effects on the main degradation phenomena is still to be assessed. This paper relates the main experimental events issued from the test together with the interpretation work coming out from code calculations. Several results were unexpected and some are of importance for safety analyses, particularly concerning cladding oxidation and fuel degradation. The test interpretation has been undertaken with the ICARE2 code, developed by IRSN “Institut de Radioprotection et de Suˆrete´ Nucle´aire”. ICARE2 is the stand alone part of the mechanistic ICARE/CATHARE code, built for safety analyses purposes. In the last code versions, models deal with the degradation and the oxidation of the B4C control rod. Already validated against QUENCH experiments, these models have been applied to the Phebus FPT-3 conditions. Calculations show that, although general trends of the test are very well reproduced, specific effects concerning the interaction of B4C with the neighbouring rods cannot be calculated. Better detailed physical analysis and prediction capabilities of calculation tools are expected to come out from the International Source Term Program, in particular from the results of the BECARRE separate-effect-tests performed at IRSN. Though the detailed modelling of such interactions has still to be improved, the Phe´bus FP series of integral experiments have proved the capability of the program to significantly improve our understanding of the source term issue and important aspects of core degradation.
Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System
Authors: Homma Shunji , Ishii Jun-Ichi , Koga Jiro , Matsumoto Shiro , Kikuchi Toshiaki , Chikazawa Takahiro , Shibata Atsuhiro
Published: September 2008

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced.
Neutron Detection with Centrifugally-Tensioned Metastable Fluid Detectors (CTMFD)
Authors: Xu Y. , Smagacz P. , Lapinskas J. , Webster J. , Shaw P. , Taleyarkhan R. P.
Published: September 2008

Tensioned metastable liquid states at room temperature were utilized to display sensitivity to impinging nuclear radiation, that manifests itself via audio-visual signals that one can see and hear. A centrifugally-tensioned metastable fluid detector (CTMFD), a diamond shaped spinning device rotating about its axis, was used to induce tension states, i.e. negative (sub-vacuum) pressures in liquids. In this device, radiation induced cavitation is audible due to liquid fracture and is visible from formed bubbles, so called hearing and seeing radiation. This type of detectors is selectively insensitive to Gamma rays and associated indication devices could be extremely simple, reliable and inexpensive. Furthermore, any liquids with large neutron interaction cross sections could be good candidates. Two liquids, isopentane and methanol, were tested with three neutron sources of Cf-252, PuBe and Pulsed Neutron Generator (PNG) under various configurations of neutron spectra and fluxes. The neutron count rates were measured using a liquid scintillation detector. The CTMFD was operated at preset values of rotating frequency and a response time was recorded when a cavitation occurred. Other parameters, including ambient temperature, ramp rate, delay time between two consecutive cavitations, were kept constant. The distance between the menisci of the liquid in the CTMFD was measured before and after each experiment. In general, the response of liquid molecules in a CTMFD varies with the neutron spectrum and flux. The response time follows an exponential trend with negative pressures for a given neutron count rate and spectra conditions. Isopentane was found to exhibit lower tension thresholds than methanol. On the other hand, methanol offered a larger tension metastability state variation for the various types of neutron sources, indicating the potential for offering significantly better energy resolution abilities for spectroscopic applications.
Qualification and Uncertainty Evaluation of a Best Estimate Loca Study of the Mochovce NPP by RELAP5/3.2-Gamma and CIAU
Authors: Kvizda Boris , Petruzzi Alessandro , D’Auria Francesco S.
Published: September 2008

The paper describes the work that has been done in cooperation between VUJE, Inc. (Slovakia) and University of Pisa (UNIPI, Italy). The aim was to evaluate Best Estimate Plus Uncertainty calculation (BEPU) of Mochovce Nuclear Power Plant WWER-440/213 (NPP) for the selected LOCA scenario utilizing the Code with the capability of Internal Assessment of Uncertainty (CIAU) developed at UNIPI. At the basis of CIAU, there is the Uncertainty Method based on Accuracy Extrapolation (UMAE) previously proposed by UNIPI. The project was considered as a full pilot BEPU application for WWER-440 type of unit in VUJE, Inc. that can be taken as a basis for utilization in licensing process later on. Project objectives are as follows: • To prove capability of available Mochovce NPP nodalization for RELAP5/3.2.2 gama to predict meaningful results for LOCA type accidents following UMAE procedure [1]. • To apply the CIAU [2, 3] to a LOCA transient selected for the Mochovce NPP and produce uncertainty bands for selected relevant parameters.
Air Ingress Accident in a High Temperature Reactor with Prismatic Fuel
Authors: Haque H. , Brinkmann G.
Published: September 2008

In this paper, the safety behavior of the new generation high temperature reactors (HTRs) with prismatic fuels during air ingress accident conditions has been investigated. These reactors conceived primarily for the production of hydrogen, are characterized by their inherent safety features with respect to passive decay heat removal through conduction, radiation and natural convection. Air ingress is an HTR specific event. The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 °C leading to reaction heat and graphite corrosion. A substantial amount of graphite burn-off can take place only if sufficient amount of air enters into the core. In order to better assess the phenomena of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is determined as a function of the break cross sections exposed above and below the core. This paper demonstrates the thermal behavior of the ANTARES reactor (operating inlet/outlet temperatures 450/850 °C) for various air flow rates with respect to graphite burn-off and maximum temperatures of fuel and bottom reflector region. It indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the pellets exposed.
Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant
Authors: Kim Sang-Nyung , Lim Sang-Gyu
Published: September 2008

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.
CFD Modeling of Thermal Effects of Nuclear Waste Vitrification Processes
Authors: Rayner Chris , Soltani Mehdi , Barringer Chris , Knight Kelly J.
Published: September 2008

The Waste Treatment Plant (WTP) at Hanford, WA will vitrify nuclear waste stored at the DOE Hanford facility. The vitrification process will take place in two large concrete buildings where the glass is poured into stainless steel canisters or containers and allowed to cool. Computational Fluid Dynamics (CFD) was used extensively to calculate the effects of the heat released by molten glass as it is poured and cooled, on the HVAC system and the building structure. CFD studies of the glass cooling in these facilities were used to predict canister temperatures, HVAC air temperatures, concrete temperatures and insulation requirements, and design temperatures for canister handling equipment and instrumentation at various stages of the process. These predictions provided critical input in the design of the HVAC system, specification of insulation, the design of canister handling equipment, and the selection of instrumentation.
Local Interfacial Structures in Horizontal Bubbly Flow with 90-Degree Bend
Authors: Kim Seungjin , Park Jung Han , Kojasoy Gunol , Kelly Joseph M.
Published: September 2008

Present study investigates the geometric effects of flow obstruction on the distribution of local two-phase flow parameters and their transport characteristics in horizontal two-phase flow. The round glass tubes of 50.3mm in inner diameter are employed as test sections, along which a 90-degee elbow is located at L/D = 206.6 from the two-phase mixture inlet. In total, 15 different flow conditions are examined within the air-water bubbly flow regime. The detailed local two-phase flow parameters are acquired by the double-sensor conductivity probe at four different axial locations. The effect of elbow is found to be evident in both the distribution of local parameters and their development. The elbow clearly promotes bubble interactions resulting in significant changes in interfacial area concentration. It is also found that the elbow-effect propagates to be more significant further downstream (L/D = 250) than immediate downstream (L/D = 225) of the elbow. Furthermore, it is shown that the elbow induces significant oscillations in the flow in both vertical and horizontal directions of the tube cross-section. Characteristic geometric effects due to the existence of elbow are also shown clearly on the axial development of one-dimensional interfacial area concentration and void fraction.
RANS Predictions in an Idealized Lower-Plenum Model
Authors: Hodson Joshua D. , Spall Robert E. , Smith Barton L.
Published: September 2008

The two-dimensional, unsteady, Reynolds-averaged Navier-Stokes equations have been solved for the flow across a row of confined cylinders with a pitch-to-diameter ratio of 1.7, a configuration which was designed to model a next generation nuclear plant lower-plenum. Four different turbulence models were used: k–ε, k–ω, v2–f, and differential Reynolds-stress transport. Comparisons with available experimental data were made for pressure losses, recirculation lengths, and mean velocity profiles. The results indicate that all models did a reasonable job of predicting the pressure loss coefficient. However, in terms of the mean velocities and recirculation length, the determination of which model performed best is not clear.
Finite Element Simulation of Dynamic Behavior of Pressurized Piping System Under Large Seismic Load
Authors: Ravi Kiran A. , Agrawal M. K. , Reddy G. R. , Ghosh A. K. , Kushwaha H. S.
Published: September 2008

The nuclear power plant piping systems are often subjected to the cyclic loading conditions due to transients, seismic or other unexpected events. During these events, the probable mode of failure for piping component is fatigue-ratcheting. Earlier the design of piping subjected to seismic excitation was based on the principle of static collapse. Later on, it was postulated that the cause of failure of piping components is fatigue ratcheting and not plastic collapse. The 1995 ASME Boiler & Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. An Analytical study is carried out to investigate the behavior of the pressurized piping system under large seismic loading. The analysis is performed using equivalent inertial forces. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. The capability of the model to simulate the ratcheting response of the piping system is of particular interest. Comparison of analysis results against test results are presented in the paper.
Predictions of Critical Heat Flux in Annular Pipes with TRACE V4.160 Code
Authors: Jasiulevicius Audrius , Macian-Juan Rafael
Published: September 2008

This paper presents the assessment of TRACE (version v4.160) against the Critical Heat Flux (CHF) experiments in annular tubes performed at the Royal Institute of Technology (KTH) in Stockholm, Sweden. The experimental database includes data for coolant mass fluxes between 250 and 2500 kg/m2s and inlet subcoolings of 10 and 40 K at a pressure of 70 bar. The work presented in this paper supplements the calculations of single round tube experiments carried out earlier and provides a broader scope of validated geometries. In addition to the Biasi and CISE-GE CHF correlations available in the code, a number of experimental points at low flow conditions are available for the annular geometry experiments, which also permitted the assessment of the Biasi/Zuber CHF correlation used in TRACE v4.160 for low flow conditions. Experiments with different axial power distribution were simulated and the effects of the axial power profile and the coolant inlet subcooling on the TRACE predictions were investigated. The results of this work show that the Biasi/Zuber correlation provides good estimation of the CHF at 70 bar, and, for the same conditions, the simulation of the annular experiments resulted in the calculation of lower CHF values compared to single-tube experiments. The analysis of the performance of the standard TRACE CHF correlations shows that the CISE-GE correlation yields critical qualities (quality at CHF) closer to the experimental values at 70 bar than the Biasi correlation for annular flow conditions. Regarding the power profile, the results of the TRACE calculations seem to be very sensitive to its shape, since, depending on the profile, different accuracies in the predictions were noted while other system conditions remained constant. The inlet coolant subcooling was also an important factor in the accuracy of TRACE CHF predictions. Thus, an increase in the inlet subcooling led to a clear improvement in the estimation of the critical quality with both Biasi and CISE-GE correlations. To complement the work, three additional CHF correlations were implemented in TRACE v4.160, namely the Bowring, Tong W-3 and Levitan-Lantsman CHF models, in order to assess the applicability of these correlations to simulate the CHF in annular tubes. The improvement of CHF predictions for low coolant mass flows (up to 1500 kg/m2s) is noted when applying Bowring CHF correlation. However, the increase in the inlet subcooling increases the error in predicted critical quality with the Bowring correlation. The Levitan-Lantsman and Tong-W-3 correlations provide results similar to the Biasi model. Therefore, the most correct CHF predictions among the investigated correlations were obtained using CISE-GE model in the standard TRAC v4.160 code.
Regulatory Endorsement Activities for ASME Nuclear Codes and Standards
Authors: West Raymond A.
Published: September 2008

The ASME Board on Nuclear Codes & Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes & Standards (NC&S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC&S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort.
Heat Transfer at Supercritical Pressures and the Onset of Deterioration
Authors: Kirillov Pavel L. , Grabezhnaya Vera A.
Published: September 2008

The comparison of the data on heat transfer at supercritical pressures (SCP) demonstrates that they have a considerable spread, which shows a complex nature of the process and a probable inaccuracy in the methods of data processing caused by a sharp change in thermophysical properties near the pseudocritical point. The recent experimental data at SCP for upward flow of water are compared with some correlations applicable to engineering analysis. The correlations for the onset of heat transfer deterioration against the experimental data were analyzed. The heat transfer deterioration in this data was far from test section inlet The generalization of data on the onset of heat transfer deterioration for various coolants (water, CO2, R12) was proposed.