The prestressed cast iron reactor pressure vessel (PCIPV)

The concept of the prestressed cast iron reactor pressure vessel (PCIPV) emerges from the utilization of cast iron in the design of radiation and thermal shields. The principles of construction are explained using a model which is at present being assembled. Salient differences between the proposed vessel concept and a prestressed concrete reactor pressure vessel (PCPV) are discussed.     

Integrated design of prestressed cast-iron pressure vessel and passive heat removal system for the reactor cell of a 200 MWth modular reactor

For a modular reactor of 200 MW thermal output an inactive after heat removal system has been designed. It consists of a prestressed cast iron pressure vessel with the surrounding reactor cell. Integrated in the cast iron profiles of the reactor cell is a redundant water cooling system based on natural convection. Air cooling towers are provided to cool the water down to ambient temperature. The cooling system covers a wide range of possible wall temperatures without significant changes in water temperature. The structures of the reactor pressure vessel and the cell, their assembly and some results of the engineering work done up to now are described in this paper.     

Development and qualification of materials for structural components for the high-temperature gas-cooled reactor

This paper summarizes the materials evaluation work which has been done for the advanced HTR projects. Essentials of the materials information needed for the design of prestressed concrete pressure vessels (PCPV), steel pressure vessels, graphite incore structures, and metallic heat components inside the pressure vessel are discussed. The main emphasis is given to the metallic high-temperature components which are exposed in the temperature regime in which all properties are sincerely temperature and time dependent. For those components and for graphitic side reflectors the methods of analysis and proof of integrity of the components during the total operation time are provided.     

Simplified distributed parameters BWR dynamic model for transient and stability analysis

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR.     

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.     

Mechanistic modeling for ring-type BWR fuel spacer design 2. Local pressure drop and narrow channel effect

The fuel spacer is one of the components of a fuel rod bundle and its role is to maintain an appropriate rod-to-rod clearance. Since the fuel spacer influences liquid film flow on fuel rods in BWR core, its specification has a strong effect on thermal hydraulics in the core such as critical power and pressure drop. Spacers have been developed through empirical modifications, so that a large amount of test data is required for optimum design of the spacer. It is one of the important subjects to develop a mechanistic model of the spacer for future design of BWR fuel bundles. The authors proposed a liquid film flow limitation model (narrow channel effect) in which the momentum balance of the liquid film phase in the near field of the spacer is considered. Modeling of local pressure drop of the spacer is required for closing the momentum equation. In this paper, the effect of spacer geometry on the local pressure drop was, therefore, investigated experimentally and analytically in a vertical circular channel using air and water as test fluids.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Basic probabilistic considerations on safety of prestressed concrete pressure vessels

The purpose of the present study is to lay the foundation for the probabilistic safety analysis pertinent to prestressed concrete pressure vessels of nuclear generating stations. We place a major emphasis on the consistency with which various aspects of structural and statistical significance are analyzed and then combined into a probabilistic formulation amenable to practical risk assessment and assurance. In particular (1) prestressed concrete pressure vessels subjected to loads of various temporal nature are considered, (2) the probabilistic algebra used in risk assessment is basically that of the second moment-first order approximation, and (3) the method thus established is applied to reliability analysis of prestressed concrete reactor vessels (PCRVs) to estimate the probability level of ultimate failure. The result of the present study indicating the probability level of the vessel reliability depends entirely on the assumptions concerning statistical aspects of loading conditions and resisting capacities of the vessel and on the approximations in relation to stress and structural analysis and ultimate failure conditions. Therefore, the resulting probability values are illustrative and comparative rather than factual at this time. The authors are solely responsible for the content of this paper which does not represent the official view of the USAEC.     

Application of IAEA TECDOC 717 to Packagings and Comparison with Reactor Vessels

Abstract

Brittle fracture evaluation is important for type B(U) packagings and packagings carrying fissile materials in conformity with IAEA Regulations. Packaging materials susceptible to brittle fracture can be evaluated by Linear Elastic Fracture Mechanics (LEFM) or other credible methods as shown in IAEA TECDOC 717. Major similarities and differences between the packagings and reactor pressure vessels are here compared in terms of the brittle fracture evaluation. Examples of brittle fracture evaluations of 100 ton class ductile cast iron (DCI) packagings (casks) under hypothetical accident conditions are discussed.     

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.     

Containment vessel design and practice

This paper attempts to review the state of the art of methods of analysis and design of the concrete containment vessels required for BWR and PWR. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis given by the author is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load.Two existing vessels in reinforced and pre-stressed concrete are examined. Substantial calculations are given in order to assess their behaviour. Both these original analyses were developed by the author and are given in the appendices. They have been fully tested on four other vesels and two models. Due to limitations of space, some of these details could not be revealed. A brief explanation is given regarding the computer programs supporting the above analysis.     

Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste.The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology.The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s.International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.     

Current and emerging pressure boundary issues — a US regulatory perspective

Concerns about pressure boundary integrity deal primarily with older plants, and establishing a basis for their continued safe operation. Pressure vessel problems stem from exposure to fast neutrons which changes the Nil-Ductility-Temperature (NDT) and the elevated temperature fracture energy of some vessels. The predicted shift in NDT has increased over the last decade as more has been learned about the effect of impurities (copper) and the synergism between nickel and copper. In PWRs this has lead to concern about excursion in which the a vessel remains at high pressure as the coolant temperature drops rapidly, that is the so-called Pressurized Thermal Shock (PTS) accident. In BWRs one cannot have PTS events, but the more rapid than expected rise in NDT due to irradiation is impacting operations.In another set of PWRs the upper shelf energy of the welds was initially low due to the use of a slag which led to many small inclusions in the weld. Radiation has lowered the Charpy fracture energy of these welds to below the 50 ft lb level at which there is concern that the vessel may undergo low energy ductile failure even if cleavage does not occur.Problems in pressure boundary piping has stemmed primarily from corrosion, that is, IGSCC in BWR recirculation piping, and steam generator tube failures in PWRs. These have made a large contribution to downtime and occupational exposure, but are not seen as significant contributors to risk. There has been some concern about the aging (loss of toughness) of cast stainless components with significant ferrite content, especially because inspection by UT is difficult.     

Analyses of pressure perturbation events in boiling water reactor

In order to design more stable and safer core configurations, experimental and theoretical studies about BWR (Boiling Water Reactor) instability have been performed to characterize the phenomenon and to predict the conditions for its occurrence. The instabilities can be caused by interdependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In this work, the RELAP5-MOD3.3 thermal-hydraulic system code and the PARCS-2.4 3D neutron kinetic code were coupled to simulate BWR transients. The pressure perturbation is considered in order to study in detail this type of transient. Two different algorithms developed at the University of Pisa were used to calculate the Decay Ratio (DR) and the natural frequency (NF) from the power oscillation signals obtained from the transient calculations. The validation of a code model set up for the Peach Bottom-2 BWR plant is performed against Low-Flow Stability Tests (LFST). The four series of Stability Tests were performed at Peach Bottom Unit 2 in 1977 at the end of cycle 2 in order to measure the reactor core stability margins at the limiting conditions used in design and safety analysis.     

Design and analysis of concrete reactor vessels: New developments, problems and trends

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance — areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of these problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep.     

The role of BWR secondary containments in severe accident mitigation: Issues and insights from recent analyses

All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by secondary containments. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident.The fission product retention capability of an intact secondary containment will depend on several factors. Recent analyses indicate that the major factors influencing secondary containment effectiveness include: the mode and location of the primary containment failure, the internal architectural design of the secondary containment, the design of the standby gas treatment system, and the ability of fire protection system sprays to remove suspended aerosols from the the secondary containment atmosphere. Each of these factors interact in a very complex manner to determine secondary containment severe accident mitigation performance.This paper presents a brief overview of US BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented.     

A New Method for Low Sensitivity Design of a Controller for a Load Following BWR Power Plant

A new method has been proposed in which the controller is approximated by two low order models, the second being the sensitivity model of the first. Thus a Kind of piecewise solution is proposed in which the system retains the same order irrespective of the number of parameters to be considered for low sensitivity design. The usefulness of the proposed technique is illustrated in the controller design for a direct cycle BWR power plant of 457 MW(thermal) with recirculation control. The mathematical model includes the reactor kinetics, hydrodynamics of the recirculation loop, pressure transients, and the load frequency control system. The response of system variables such as frequency, neutron power, and reactor pressure are plotted with the low sensitivitty controller. The sensitivity function of frequency has been plotted using the conventional and proposed low sensitivity controller for 20 per cent variation in parameter values. The method is specifically recommended for controller design of large size systems.     

Evaluation of Dynamic Loads Induced by Chugging

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain.

As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation.

The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.     

Protection of the containment of the twin BWR KRB II plant against overpressure failure during severe accidents

The 2 × 1310 MWel plant KRB II as the newest design BWR plant in Germany (with prestressed concrete containment) had to go, with respect to the other BWRs, slightly different ways in solving the problem of containment overpressure protection during severe accidents. The basic concept of the plant and the special boundary conditions of the twin unit concept required modified solutions for the realized containment filtered venting system and the measures against energy release due to HZ reactions.The paper reports on the objectives, design, arrangement and installation of the equipment provided hereto. Special emphasis is layed upon the operation experience with the equipment installed. Furthermore, an outlook is given on possible complementary installations such as HZ-igniters and -recombiners and containment atmosphere monitoring and sampling.     

Economic analysis of hydride fueled BWR

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 × 10 hydride fuel bundles instead of 10 × 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the “per volume base” fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.