bluuebunny/crossref_metadata_2025 huggingface dataset:
Pebble Bed Boiling Water Reactor Concept With Superheated Steam
Authors: Tsiklauri G. , Newman D. , Meriwether G. , Korolev V.
Published: March 2009

An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO2 with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 °C elevates the steam cycle to high thermal efficiency of 42%.
The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture
Authors: Gu¨ntay Salih , Dehbi Abdel , Suckow Detlef , Birchley Jon
Published: March 2009

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes’ models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an unisolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (AeRosol Trapping In a Steam generaTor), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007.
Progress in Decommissioning of Ignalina NPP Unit 1
Authors: Ancius Darius , Krenevicius Rimantas , Kutas Saulius , Chouha Michel
Published: March 2009

The aim of the paper is to present the Lithuanian legal framework regarding the nuclear safety in Decommissioning and Waste Management, and the progress in the Decommissioning Programme of the unit 1 of Ignalina Nuclear Power Plant (INPP). INPP is the only nuclear plant in Lithuania. It comprises two RBMK-1500 reactors. After Lithuania has restored its independence, responsibility for Ignalina NPP was transferred to the Republic of Lithuania. To ensure the control of the Nuclear Safety in Lithuania, The State Nuclear Power Safety Inspectorate (VATESI) was created on 18 October 1991, by a resolution of the Lithuanian Government. Significant work has been performed over the last decade, aiming at upgrading the safety level of the Ignalina NPP with reference to the International standards. On 5 October 1999 the Seimas (Parliament) adopted the National Energy Strategy: • It has been decided that unit 1 of Ignalina NPP will be closed down before 2005, • The conditions and precise final date of the decommissioning of Unit 2 will be stated in the updated National Energy strategy in 2004. On 20–21 June 2000, the International Donors’ Conference for the Decommissioning of Ignalina NPP took place in Vilnius. More than 200 Millions Euro were pledged of which 165 M€ funded directly from the European Union’s budget, as financial support to the Decommissioning projects. The Decommissioning Program encompasses legal, organizational, financial and technical means including the social and economical impacts in the region of Ignalina. The Program is financed from International Support Fund, State budget, National Decommissioning Fund of Ignalina NPP and other funds. Decommissioning of Ignalina NPP is subject to VATESI license according to the Law on Nuclear Energy. The Government established the licensing procedure in the so-called “Procedure for licensing of Nuclear Activities”; and the document “General Requirements for Decommissioning of the Ignalina NPP” has been issued by VATESI. A very important issue is the technical support to VATESI and the Lithuanian TSO’s (Technical Support Organisations) in their activities within the licensing process related to the Decommissioning of INPP. This includes regulatory assistance in the preparation of decommissioning and radioactive waste management regulatory documents, and technical assistance in the review of the safety case presented by the operator. The Institute for Radioprotection and Nuclear Safety (IRSN, France) and the French Nuclear Safety Authority (DSIN) as well as Swedish International Project (SIP) are providing their support to VATESI in these areas.
Human Reliability and the Current Dilemma in Human-Machine Interface Design Strategies
Authors: Passalacqua Roberto , Yamada Fumiaki
Published: March 2009

Since human error dominates the probability of failures of still-existing human-requiring systems (as the Monju reactor), the human-machine interface needs to be improved. Several rationales may lead to the conclusion that “humans” should limit themselves to monitor the “machine”. For example, this is the trend in the aviation industry: newest aircrafts are designed to be able to return to a safe state by the use of control systems, which do not need human intervention. Thus, the dilemma whether we really need operators (for example in the nuclear industry) might arise. However, social-technical approaches in recent human error analyses are pointing out the so-called “organizational errors” and the importance of a human-machine interface harmonization. Typically plant’s operators are a “redundant” safety system with a much lower reliability (than the machine): organizational factors and harmonization requirements suggest designing the human-machine interface in a way that allows improvement of operator’s reliability. In addition, taxonomy studies of accident databases have also proved that operators’ training should promote processes of decision-making. This is accomplished in the latest trends of PSA technology by introducing the concept of a “Safety Monitor” that is a computer-based tool that uses a level 1 PSA model of the plant. Operators and maintenance schedulers of the Monju FBR will be able to perform real-time estimations of the plant risk level. The main benefits are risk awareness and improvements in decision-making by operators. Also scheduled maintenance can be approached in a more rational (safe and economic) way.
An Approach of Uncertainty Evaluation for Thermal-Hydraulic Analysis
Authors: Ogura Katsunori , Ninokata Hisashi
Published: March 2009

An approach to evaluate uncertainty systematically for thermal-hydraulic analysis programs is demonstrated. The approach is applied to the Peach Bottom Unit 2 Turbine Trip 2 Benchmark and is validated.
Assessment of Aging of Nuclear Power Plant Civil Structures
Authors: Naus D. J. , Ellingwood B. R. , Graves H. L.
Published: March 2009

Research is being conducted by ORNL for the USNRC to address aging of civil structures in light-water reactor plants. The importance and operating experience of nuclear power plant (NPP) civil structures is reviewed. Factors that can lead to age-related degradation of reinforced concrete structures and containment metallic pressure boundaries (i.e., steel containments and liners of reinforced concrete containments) are identified and their manifestations described. Background information and data for improving and developing methods to assess the effects of age-related degradation on structural performance are provided. Techniques for detection of degradation are reviewed and research related to development of methods for inspection of inaccessible regions of the containment pressure boundary presented. Application of structural reliability analysis methods to develop condition assessment tools and guidelines is described.
A Study of Steam Generator Aging Management at Selected U.S. Nuclear Plants
Authors: Lin James C.
Published: March 2009

This paper presents insights obtained from a study of the steam generator aging management at selected U.S. nuclear plants. The primary objective of this study was to conduct an in-depth review of the current steam generator (SG) management practices at selected U.S. nuclear plants. The plants chosen for review were those with similar SG design features as contained in Westinghouse Model F series SGs with thermally treated tubes. This study focuses on several major areas including inspection and integrity assessment, preventive measures (e.g., secondary-side water chemistry control and cleaning) to minimize further damage, monitoring and mitigation, and long-term steam generator replacement strategy.
The Effect of the Hydrogen to Heavy Metal Ratio (H/HM) on Reactivity and Discharge Isotopics of Homogeneous Thoria-Urania Fuel
Authors: Weaver Kevan D. , Herring J. Stephen
Published: March 2009

Calculations were performed using MOCUP, which includes the use of MCNP for neutron transport and ORIGEN for depletion. The MOCUP calculations were done using a unit cell (pin cell) model, where the ThO2 varied from 65–75wt% and the UO2 varied from 25–35wt%. The fission products and actinides being tracked in the calculations account for >97% of the parasitic captures in the fuel. The fuel pin was surrounded by four reflecting planes, where typical parameters were used for a 17×17 PWR assembly. The hydrogen to heavy metal ratio (H/HM) was varied by increasing or decreasing the water density in the cell. The results show that the drier lattices have insufficient reactivity due to the limited enrichment of the uranium. However, a slightly wetter lattice will increase the reactivity-limited burnup by 26% for the 25% UO2 – 75% ThO2, and 19% for the 35% UO2 – 65% ThO2 as compared to the standard coolant density. This is appears to be consistent with similar studies done with all-uranium lattices, where advantages are gained by hardening or further softening the neutron spectrum. Although some advantage is gained by softening the spectrum, the same can be said of all-uranium fueled cores. The spectral changes and, to a lesser extent, competing resonances between the 238U and bred-in 233U appear to hamper advantages in the conversion of thorium in homogeneous fuel that might otherwise be gained by shifting the neutron spectrum. Physically separating the uranium and thorium (e.g., in micro-heterogeneous and seed-and-blanket arrangements) have been shown alleviate this problem. A change in moderation may further enhance the reactivity-limited burnup of these lattices, and will be the focus of future work.
Design of Compact and Full-Size Simplified Boiling Water Reactors With Fully Passive Safety Systems
Authors: Ishii M. , Revankar S. T. , Xu Y.
Published: March 2009

Scientific designs of two next-generation simplified boiling water reactors (SBWRs) namely, a compact modular 200 MWe SBWR and a full-size 1200-MWe SBWR have been developed. The design involved identification of principal design criteria dictated by the safe operation of the reactor, identification of coolant requirements, and the design of the engineered safety and emergency cooling systems based on passive systems. A detailed scaling analysis was performed. The results of the scaling study were used in the performance of the integral tests and data analysis. The scaling analysis identified key thermalhydraulics parameters that govern flow phenomena in SBWRs. The analysis was based on the three-level scaling approach.
Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments
Authors: Jasiulevicius Audrius , Sehgal Bal Raj
Published: March 2009

The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchiatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR’s), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were given in the Experiment report [1] as parameters, to be obtained using static calculations, i.e. the reported results were already processed numerically using the facility equipment, e.g. the reactimeter. The reported measurement errors consist only of instrumentation errors, i.e. measurement method errors and the influence from the space–time effects were not included in the error evaluation.
VIPRE02 Code Assesment for CHF and Post-CHF Heat Transfer Modes in Long Tubes
Authors: Jasiulevicius Audrius , Sehgal Bal Raj
Published: March 2009

The paper presents validation of the CORETRAN-01 code thermal–hydraulic module VIPRE02 against experimental data, obtained at experimental facilities at EREC E108 (Russia) and RIT (Sweden). The validation was carried out in order to assess the VIPRE02 code predictions for CHF occurrence and post–CHF heat transfer in 7.0 m long tubes. The lengths of the test sections in the experiments correspond to the heated length of the RBMK reactor fuel assembly. The CHF and post-CHF (transition and film boiling) correlations, included in the standard version of VIPRE02 code package were tested. The validation results as well as VIPRE model and code modification description are the subject of this paper.
Conceptual Analysis of Fission Fragment Magnetic Collimator Reactors
Authors: Tsvetkov Pavel V. , Parish Theodore A.
Published: March 2009

As part of the current research work within the US DOE NERI Direct Electricity Conversion (DEC) Project on methods for utilizing direct electricity conversion in nuclear reactors, a detailed study of a Fission Fragment Magnetic Collimator Reactor (FFMCR) has been performed. The FFMCR concept is an advanced DEC system that combines advantageous design solutions proposed for application in both fission and fusion reactors. The present study was focused on determining the electrical efficiency and other important operational aspects of the FFMCR concept. In principle, acceptable characteristics have been demonstrated, and results obtained are presented in the paper. Technological visibility of the FFMCR concept and required further design development are discussed. Preliminary characteristics of the promising design are outlined.
Transverse Vibrations of Double Bellows Expansion Joint Restrained Against Rotation
Authors: Kameswara Rao C. , Radhakrishna M.
Published: March 2009

The paper presents the results of investigation of transverse vibrations of double bellows expansion joint restrained against rotation on either end. A theoretical model is developed based on the Bernoulli-Euler beam theory and includes added mass of the fluid flowing inside the pipe-bellow-pipe assembly. Neglecting effect of shear, an exact frequency equation is derived for the transverse vibrations of double bellows expansion joint including the effects of end elastic restraints against rotation. Numerical results are presented for an example bellow showing the effects of variation of elastic restraints and internal pressure on the first two modes of vibration.
Experimental and Analytical Studies on Penetration Integrity of the Reactor Vessel Under External Vessel Cooling
Authors: Park Rae-Joon , Kang Kyoung-Ho , Kim Jong-Tae , Koo Kil-Mo , Kim Sang-Baik , Lee Ki-Young
Published: March 2009

Experimental and analytical studies on the penetration integrity of the reactor vessel in the APR (Advanced Power Reactor) 1400 have been performed under the condition of external vessel cooling in a severe accident. The objective of this study is to estimate failure or non-failure of the penetration including the ICI (In-Core Instrumentation) nozzle and the thimble tube. Five tests in conditions with and without external vessel cooling have been performed to estimate the effects of system, corium mass, and vessel geometry using alumina (Al2O3) melt as a simulant. The test results have been evaluated using the LILAC (Lower head IntegraL Analysis computer Code). The tests results have shown that penetration in the no external vessel cooling case is more damaged than that in the external vessel cooling case. An increase in system pressure from 0.9 MPa to 1.5 MPa was not effective on penetration damage, but an increase in corium mass from 40 kg to 60 kg and a vessel geometry change to flat plate with curvature were effective. The LILAC results are very similar to the test results on the ablation depth in the weld. It is concluded that external vessel cooling is a very effective means for maintaining penetration integrity.
Preliminary Probabilistic Safety Assessment of the IRIS Plant
Authors: Mizuno Yuko O. , Ogura Katsunori , Ninokata Hisashi , Conway Lawrence E.
Published: March 2009

A preliminary level-1 probabilistic safety assessment of the IRIS plant has been performed. The first focus is on five internal initiating events, such as primary system break (loss-of-coolant accident and steam generator tube rupture) and transients (secondary system line break and loss-of-off-site power). In this study, the event tree for each initiating event was developed and the fault tree analysis of the event tree headings was carried out. In particular, since one of the IRIS safety systems, the passive emergency heat removal system, is unique to the IRIS plant and its reliability is key to the core damage frequency evaluation, it received more extensive fault-tree development. Finally the dominant sequences that lead to severe accidents and the failures in the main and support systems are identified.
Experimental Investigation Into Thermal Siphon Used as an Intermediate Circuit of an Integrated Cooling System Reactor
Authors: Adamovich L. A. , Gabaraev B. A. , Solovjev S. L. , Shpansky S. B.
Published: March 2009

In the paper the results of study in heat transfer capacity of the themosyphon mock-up which is considered as an intermediate circuit of the reactor under design, are presented. The mock-up design, the test rig and the experimental results are described. It is shown that the simplest mathematical model describes the processes of power transfer by the thermosyphon under certain conditions.
Computed Radiography: An Innovative Inspection Technique
Authors: Klein William A. , Councill Donald L.
Published: March 2009

Florida Power & Light Company’s (FPL) Nuclear Division combined two diverse technologies to create an innovative inspection technique, Computed Radiography, that improves personnel safety and unit reliability while reducing inspection costs. This technique was pioneered in the medical field and applied in the Nuclear Division initially to detect piping degradation due to flow-accelerated corrosion. Component degradation can be detected by this additional technique. This approach permits FPL to reduce inspection costs, perform on line examinations (no generation curtailment), and to maintain or improve both personnel safety and unit reliability. Computed Radiography is a very versatile tool capable of other uses: • improving the external corrosion program by permitting inspections underneath insulation, and • diagnosing system and component problems such as valve positions, without the need to shutdown or disassemble the component.
Pipe Decontamination Involving “String-Foam Circulation”
Authors: Turchet J. P. , Fournel B. , Estienne G.
Published: March 2009

Foam applications number for nuclear decontamination purposes has recently increased. The major advantage of foam decontamination is the reduction of secondary liquid wastes volumes. Among foam applications, we focus on foam circulation in contaminated equipements. Dynamic properties of the system ensures an homogeneous and rapid effect of the foambed-drifted chemical reagents present in the liquid phase. This paper describes a new approach of foam decontamination for pipes. It is based on an alternated air and foam injections. We called it “string-foam circulation”. A further reduction of liquid wastes is achieved compared to continuous foam. Secondly, total pressure loss along the pipe is controlled by the total foam length in the pipe. It is thus possible to clean longer pipes keeping the pressure under atmospheric pressure value. This ensures the non dispersion of contamination. This study describes experimental results obtained with a neutral foam as well with an acid foam on a 130 m long loop. Finally, the decontamination of a 44 meters pipe is presented.
Decontamination of Phebus Experimental Target Chamber Using Sprayed Foam
Authors: Fournel B. , Angot S. , Joyer P.
Published: March 2009

Phebus facility was designed in the eighties to lead inertial confinement fusion experiments on deuterium/ tritium targets. It has been decommissioned in 1999. Phebus chamber is an aluminum alloy (5086) sphere of 2,3 m diameter with an inner volume of about 8 m3. The thickness of the chamber is about 12 cm. A large number of openings (around 230) were designed on its surface, for diagnostics implementation during experiments (figure 1).
Investigations on the Melt Gate Ablation by Ex-Vessel Core Melts in the KAPOOL Experiments
Authors: Eppinger Beatrix , Schmidt-Stiefel Silke , Tromm Walter
Published: March 2009

In future Light Water Reactors (LWR) containment failure should be prevented even for very unlikely core meltdown sequences with reactor pressure vessel (RPV) failure. In the case of such a postulated core meltdown accident in a future LWR the ex-vessel melt shall be retained and cooled in a special compartment inside the containment to exclude significant radioactive release to the environment. In such a case, a gate has to be designed to allow the melt release from the reactor cavity into the compartment. A series of transient experiments has been performed to investigate the melt gate ablation using iron and alumina melts as a simulant for the corium melt. The results of the KAPOOL tests are analyzed with the HEATING5 code in order to evaluate realistic cases of internally heated corium melts and melt gates with the same theoretical tool.
Study on Analytical Prediction of Forced Convective CHF in the Wide Range of Quality
Authors: Kodama S. , Kataoka I.
Published: March 2009

For the purpose of predicting CHF for a wide range in quality, we developed the analytical CHF prediction method. Two-phase flow analysis code based on multi-fluid model was developed and, by using typical dryout model and DNB model, CHF predictions for a wide range in quality were carried out, firstly. The dryout model and DNB model gave good predictions for high quality conditions and low quality conditions respectively. The boundary between high quality and low quality seemed to be about 0.1 to 0.2, which corresponds to the annular mist flow transition criterion. Based on this result, secondly, we carried out CHF predictions by using dryout model or DNB model selectively depending on the flow regime and got agreement to some extent with CHF data for a wide range in quality.
Independent Qualification of the CIAU Tool Based on the Uncertainty Estimate in the Prediction of Angra 1 NPP Inadvertent Load Rejection Transient
Authors: Borges Ronaldo C. , D’Auria Francesco , Alvim Antonio Carlos M.
Published: March 2009

The Code with - the capability of - Internal Assessment of Uncertainty (CIAU) is a tool proposed by the “Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP)” of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, “Comissa˜o Nacional de Energia Nuclear”, contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in the prediction of system code calculations for light water reactors and is internally coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the Angra 1 NPP transient RES-11. 99 originated by an inadvertent complete load rejection that caused the reactor scram when the unit was operating at 99% of nominal power. The current limitation of the ‘error’ database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated.
Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS
Authors: Ambrosek Richard G. , Pedersen Robert C. , Maple Amanda
Published: March 2009

Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axi-symmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique.
Quantifying the Metrics That Characterize Safety Culture of Three Engineered Systems
Authors: Tucker Julie , Ernesti Mary , Tokuhiro Akira
Published: March 2009

With potential energy shortages and increasing electricity demand, the nuclear energy option is being reconsidered in the United States. Public opinion will have a considerable voice in policy decisions that will “roadmap” the future of nuclear energy in this country. This report is an extension of the last author’s work on the “safety culture” associated with three engineered systems (automobiles, commercial airplanes, and nuclear power plants) in Japan and the United States. Safety culture, in brief is defined as a specifically developed culture based on societal and individual interpretations of the balance of real, perceived, and imagined risks versus the benefits drawn from utilizing a given engineered systems. The method of analysis is a modified scale analysis, with two fundamental eigenmetrics, time- (τ) and number-scales (N) that describe both engineered systems and human factors. The scale analysis approach is appropriate because human perception of risk, perception of benefit and level of (technological) acceptance are inherently subjective, therefore “fuzzy” and rarely quantifiable in exact magnitude. Perception of risk, expressed in terms of the psychometric factors “dread risk” and “unknown risk”, contains both time- and number-scale elements. Various engineering system accidents with fatalities, reported by mass media are characterized by τ and N, and are presented in this work using the scale analysis method. We contend that level of acceptance infers a perception of benefit at least two orders larger magnitude than perception of risk. The “amplification” influence of mass media is also deduced as being 100- to 1000-fold the actual number of fatalities/serious injuries in a nuclear-related accident.
Progress in Development of a Converging Beam Neutron Source for Driving a Sub-Critical Fission Reactor
Authors: Miley G. H. , Momota H. , Shaban Y. , Hora H.
Published: March 2009

Several laboratories are studying the possibility of a fission reactor system based on driving a sub-critical assembly using an accelerator-spallation target neutron source. The objective is to effectively eliminate possible criticality and meltdown accidents, increasing plant safety. However, one disadvantage is the large cost projected for the accelerator-driven source. In an ICONE-8 paper we proposed to overcome this problem by use of a Converging Beam Neutron Source (CBNS) to produce 14-MeV D-T fusion neutrons to drive the sub-critical core. [1] The CBNS is analogous to an accelerator-plasma target device with built-in re-circulation of the ions. It offers the unique advantage of being small enough to allow insertion of multiple “modular” units in fuel channel locations (cf. the large single target used in accelerator-drive designs). As proposed in an ICONE-9 paper, a first important step in development of such systems might be use in low power research reactors. [2] This reduces the neutron source strength requirement to a level only slightly above that obtained with present IEC experiments. Still, a key step for CBNS development is to increase the neutron production efficiency obtained in previous small-scale experiments. To do this we have recently developed a unique RF-driven ion source so that the ion production region can be separated from the main CBNS chamber. This has the combined advantages of allowing ion production at relatively high pressure, while the CBNS chamber can be pumped to ultra-low pressure. Initial experiments with this arrangement are presented here and it is shown that a very favorable scaling to the yields required for research reactor operation are predicted.
Effects of Dynamic Strain-Aging and Cyclic Loading on Fracture Behavior of Ferritic Steels
Authors: Murty K. Linga , Seok Chang-Sung
Published: March 2009

Ferritic steels commonly used for pressure vessels and reactor supports in light water reactors (LWRs) exhibit dynamic strain aging (DSA) resulting in decreased ductility and toughness. In addition, recent work indicated decreased toughness during reverse-cyclic loading that has implications on reliability of these structures under seismic loading conditions. We summarize some of our recent work on these aspects along with synergistic effects, of interstitial impurity atoms (IIAs) and radiation induced point defects, that result in interesting beneficial effects of radiation exposure at appropriate temperature and strain-rate conditions. Radiation-defect interactions were investigated on pure iron, Si-killed mild steel, A533B, A516, A588 and other reactor support and vessel steels. In all cases, DSA is seen to result in decreased ductility accompanied by increased work-hardening parameter. In addition to mechanical property tests, fracture toughness is investigated on both A533B and A516 steels. While dips in fracture toughness are observed in A533B steel in the DSA region, A516 steel exhibited at best a plateau. The reasons could lie in the applied strain-rates; while J1c tests were performed on A533B steel using 3-point bend tests on Charpy type specimens, CT specimens were used for A516 steel. However, tensile and 3-point bend tests on similar grade A516 steel of different vintage did exhibit distinct drop in the energy to fracture. Load-displacement curves during J1c tests on CT specimens did show load drops in the DSA regime. The effect of load ratio (R) on J versus load-line displacement curves for A516 steel is investigated from +1 to −1 at a fixed normalized incremental plastic displacement of 0.1 (R = 1 corresponds to monotonic loading). We note that J-values are significantly reduced with decreasing load ratio. The work-hardening characteristics on the fracture surfaces were studied following monotonic and cyclic loading fracture tests along with the stress-field analyses. From the hardness and the ball-indentation tests, it was shown that decreased load ratio (R) leads to more strain hardening at the crack tip resulting in decreased fracture toughness. From the stress field analysis near the crack tip of a compact tension fracture toughness test specimen, a cycle of tensile and compressive loads is seen to result in tensile residual stresses (which did not exist at the crack tip before). These results are important to evaluations of flawed-structures under seismic loading conditions, i.e. Leak-Before-Break (LBB) and in-service flaw evaluation criteria where seismic loading is addressed. In addition, studies on fast vs total (thermal+fast) neutron spectra revealed unexpected results due to the influence of radiation exposure on source hardening component of the yield stress; grain-size of pure iron plays a significant role in these effects.
Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management
Authors: Kang K. H. , Park R. J. , Koo K. M. , Kim S. B. , Kim H. D.
Published: March 2009

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.
A Supercritical CO2 Gas Turbine Power Cycle for Next-Generation Nuclear Reactors
Authors: Dostal Vaclav , Driscoll Michael J. , Hejzlar Pavel , Todreas Neil E.
Published: March 2009

Although proposed more than 35 years ago, the use of supercritical CO2 as the working fluid in a closed circuit Brayton cycle has so far not been implemented in practice. Industrial experience in several other relevant applications has improved prospects, and its good efficiency at modest temperatures (e.g., ∼45% at 550°C) make this cycle attractive for a variety of advanced nuclear reactor concepts. The version described here is for a gas-cooled, modular fast reactor. In the proposed gas-cooled fast breeder reactor design of present interest, CO2 is also especially attractive because it allows the use of metal fuel and core structures. The principal advantage of a supercritical CO2 Brayton cycle is its reduced compression work compared to an ideal gas such as helium: about 15% of gross power turbine output vs. 40% or so. This also permits the simplification of use of a single compressor stage without intercooling. The requisite high pressure (∼20 MPa) also has the benefit of more compact heat exchangers and turbines. Finally, CO2 requires significantly fewer turbine stages than He, its principal competitor for nuclear gas turbine service. One disadvantage of CO2 in a direct cycle application is the production of N-16, which will require turbine plant shielding (albeit much less than in a BWR). The cycle efficiency is also very sensitive to recuperator effectiveness and compressor inlet temperature. It was found necessary to split the recuperator into separate high- and low-temperature components, and to employ intermediate recompression, to avoid having a pinch-point in the cold end of the recuperator. Over the past several decades developments have taken place that make the acceptance of supercritical CO2 systems more likely: supercritical CO2 pipelines are in use in the western US in oil-recovery operations; 14 advanced gas-cooled reactors (AGR) are employed in the UK at CO2 temperatures up to 650°C; and utilities now have experience with Rankine cycle power plants at pressures as high as 25 MPa. Furthermore, CO2 is the subject of R&D as the working fluid in schemes to sequester CO2 from fossil fuel combustion and for refrigeration service as a replacement for CFCs.
Managing the Cost of Plant Piping System Leakage
Authors: Jenco John M. , Keck Donna R. , Johnson Gary L.
Published: March 2009

Recent studies have shown that the average annual cost impact of external piping system leakage on commercial nuclear plant operations and maintenance can easily range into the millions of dollars for each reactor unit. Evidence suggests that this significant O&M cost reduction opportunity has largely been overlooked, due to the number of diverse line items and budget areas affected. Results released last year from an EPRI pilot study of more than a dozen reactor units at seven plant sites operated by multiple utilities found that the average annual cost impact was indeed around $1.6 million per year per unit. Subsequent field experience has also demonstrated that an effective fluid leak management program can substantially reduce these costs within the first three years of implementation. This paper presents the general cost impact research results from various studies, outlines key elements of an effective plant fluid leak management program, discusses important implementation issues, and presents results from case studies covering different utility approaches to developing and implementing an effective fluid leak management program. Actual cost data will be included where appropriate.
The Redistribution of Load in Bolted Gasketed Joints Subjected to Steady State Thermal Loading
Authors: Nechache Akli , Bouzid Abdel-Hakim
Published: March 2009

Pressure vessel joints operating at high temperature are often very difficult to seal. The existing flange design methods do not address thermal effects other than the variation of flange material mechanical properties with temperature. It is possible to include the effects of temperature loading in joint analysis, however, presently very few guidelines exist for this type of analysis. This paper outlines the theoretical analysis used for the determination of the steady state operating temperature and the induced loads in flange joints. It details the theoretical equations necessary to predict the temperature and the redistribution of load due to the thermal expansion of the joint components for the case of a pair flange and the case of a flange with a blind-cover. The results from the theoretical models are verified by comparison to finite element results.
ADAM: An Accident Diagnostic, Analysis and Management System — Applications to Severe Accident Simulation and Management
Authors: Zavisca M. J. , Khatib-Rahbar M. , Esmaili H. , Schulz R.
Published: March 2009

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM’s severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of “alarms” are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management.
Three-Dimensional Characteristics of Post-CHF Behaviour Within a Rod Bundle for Loss-of-Flow Simulation: Experimental and Three-Fluid Porous Media Numerical Analyses
Authors: Stosic Zoran V. , Stevanovic Vladimir D. , Iguchi Tadashi
Published: March 2009

The investigation of three-dimensional transient propagations of dry-out fronts within a nuclear fuel rod bundle is performed, based on experimental and numerical simulations. The spreading of Critical Heat Flux (CHF) fronts across a bundle, caused by sudden decrease of coolant mass flow rate followed by delayed gradual decrease of power generation is predicted, and the locus of dry patches is shown. Simultaneous occurrence of CHF and rewet multi-fronts in here-analysed flow transient has not been detected so obvious as in power transient, previously analysed. Due to a possible building of a vapour zone, the CHF front spatial propagation has to be carefully analysed in transient conditions.
Influence of the Fin on Two-Dimensional Characteristics of Dispersed Flow With Wall Liquid Film in the Vicinity of Obstacle
Authors: Stosic Zoran V. , Stevanovic Vladimir D. , Serizawa Akimi
Published: March 2009

Spacers have positive effects on the heat transfer enhancement and critical heat flux (CHF) increase downstream of their location in the boiling channel. These effects are further increased by the inclusion of the fin on the spacer rear edge. Numerical simulation of a separation in a high void gas phase and dispersed droplets flow around a spacer, with a liquid film flowing on the wall, is performed. Mechanisms leading to the CHF increase due to the two-phase flow separation and liquid film thickening downstream the spacer are demonstrated. Numerical simulations of gas phase, entrained droplets and wall liquid film flows were performed with the three-fluid model and with the application of the high order numerical scheme for the liquid film surface interface tracking. Predicted is a separation of gas and entrained droplets streams around the spacer without and with a fin inclined 30 and 60 degrees to the wall, as well as a change of wall liquid film thickness in the vicinity of spacer. Results of liquid film dynamic behaviour are compared with the recently obtained experimental results. Multi-dimensional characteristics of surface waves on the liquid film were measured with newly developed ultrasonic transmission technique in a 3×3 rod bundle test section with air-water flow under atmospheric conditions. Obtained numerical results are in good agreement with experimental observations. The presented investigation gives insight into the complex mechanisms of separated two-phase flow with wall liquid film around the spacer and support thermal-hydraulic design and optimisation of flow obstacles in various thermal equipments.