Primary System Components Research Articles

Thermal transient test and strength evaluation of a thick cylinder model made of Mod.9Cr-1Mo steel

Mod.9Cr-1Mo steel is a candidate material for the primary and secondary heat transport system components of the Japan sodium-cooled fast reactor. However, there is little evidence to support the structural integrity of components made of Mod.9Cr-1Mo steel in a practical environment. To verify the failure mode and assess creep–fatigue damage, a thick cylinder test model made of Mod.9Cr-1Mo steel was subjected to 1873 cycles of accelerated thermal transient loading using a large-scale sodium loop through which liquid sodium at 600°C and 250°C flowed repeatedly, with the period of each transient being 2h and 1h, respectively. After completion of the test, the test model was inspected using liquid penetrant testing. Observations using a scanning electron microscope and hardness testing were then performed to characterize creep–fatigue damage in the structural model subjected to cyclic thermal transient loading in a sodium environment. Heat transfer analysis based on the measured temperature data and elastic and inelastic thermal stress analyses using finite element analysis were performed to evaluate the relationship between creep–fatigue damage and the observed crack conditions. Finally, the characteristics of creep–fatigue damage under cyclic thermal loading in a sodium environment and the evaluation methods were considered.

Verification of the Estimation Methods of Strain Range in Notched Specimens Made of Mod.9Cr-1Mo Steel

Several methods of estimating strain range at a structural discontinuity have been developed in order to assess component reliability. In a component design at elevated temperature, estimation of strain range is required to evaluate the fatigue and creep-fatigue damage. Therefore, estimation of strain range is one of the most important issues when evaluating the integrity of a component during its lifetimes. To verify the methods of estimating strain range for discontinuous structures, low cycle fatigue tests were carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo steel, because it is a candidate material for a primary and secondary heat transport system components of Japan Sodium-cooled Fast Reactor (JSFR). Displacement control fatigue tests and thermal fatigue tests were performed by ordinary uniaxial push–pull test machine and equipment generating the thermal gradient in the notched plate by induction heating. Several notch radii were employed to vary the stress concentration level in both kinds of tests. Crack initiation and propagation process during the tests were observed by a digital microscope and the replica method to define the failure cycles. Elastic and inelastic finite element analyses were also performed to estimate strain range for predicting fatigue life. Then, these predictions were compared with the test results. Several methods such as stress redistribution locus (SRL) method, simple elastic follow-up (SEF) method, Neuber's law, and the procedures employed by elevated temperature design codes were applied. Through these comparisons, the applicability and conservativeness of these strain range estimation methods, which is the basis of the fatigue and creep-fatigue life prediction, are discussed.

태양광 추진 항공기의 초기 사이징을 위한 에너지 균형 및 구속조건 연구

태양광 추진 항공기는 감시, 지구 모니터링, 통신 등에 대기 오염 없이 그리고 가까운 장래에 인공위성과의 가격 경쟁력까지 갖추게 될 수 있어 미래의 고고도 장기체공 임무수행을 위해 더욱더 그 중요성이 증대되고 있다. 그러나 전통적인 항공기 사이징 방법들은 태양광 추진 항공기에 바로 적용될 수 없다. 본 연구에서는 다양한 동력 시스템 구성품들이 태양광 추진 장기 체공 항공기의 사이징에 어떤 영향을 미치는지를 파악하기 위하여 에너지 균형 및 구속조건 연구를 수행하였다. 본 연구에서는 동력 생성과 연료전지의 재생 에너지 저장을 위한 광전지 모듈을 동력 시스템 구성품으로 고려하였다. 또한 본 연구 결과를 검증하기 위해 고고도 무인기에 이 새로운 사이징 기법을 적용하여 결과를 제시하였다. Solar powered aircraft are becoming more and more interesting for future long endurance missions at hight altitudes, because they could provide surveillance, earth monitoring, telecommunications, etc. without any atmospheric pollution and hopefully in the near future with competitive costs compared with satellites. However, traditional aircraft sizing methods currently employed in the conceptual design phase are not immediately applicable to solar powered aircraft. Hence, energy balance and constraint analyses were performed to determine how various power system components effect the sizing of a solar powered long endurance aircraft. The primary power system components considered in this study were photovoltaic (PV) modules for power generation and regenerative fuel cells for energy storage. To verify current research results, these new sizing methods were applied to HALE aircraft and results were presented.

Thermal analysis and design of the aerial camera’s primary optical system components

Abstract To meet the thermal index requirements of an aerial camera with long focal length, three kinds of heat transfer modes including conduction, convection and radiation were analyzed. The measures were designed for the camera’s primary optical system components based on this analysis. The design principles and implementing regulations were introduced. The camera’s finite element thermal model has been established. The results of model’s temperature distribution have been obtained after the calculation according to the boundary conditions. A heat balance test has got done for the camera. The results show that both axial temperature difference and radial temperature difference are less than 5 °C, which meets the camera’s thermal control indexes.

Iris small break loca phenomena identification and ranking table (PIRT)

The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components – reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms – are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design™ approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist. While the IRIS Safety-by-Design™ approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts. To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally recognized experience in reactor safety analysis, and were not biased by program preconceptions internal to the IRIS program. The SBLOCA PIRT Panel concluded that continued experimental data and analytical tool development in the following areas, in decreasing level of significance, are perceived as important with respect to satisfying the safety analysis and licensing objectives of the IRIS program: (1) steam generator; (2) pressure suppression system, containment dry well and their interactions; (3) emergency heat removal system; (4) core, long-term gravity makeup system, automatic depressurization system, and pressurizer; (5) direct vessel injection system and reactor vessel cavity.

Gamma dose from activation of internal shields in IRIS reactor

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Modular integral water reactors: The ultimate in safety and simplicity

The integral water reactor design (i.e. a design where all primary system components—steam generators, pumps, pressurizer and even control rod drive mechanisms-are within the reactor vessel) is presented here. A background discussion of how integral designs were introduced in the 1990s is provided, followed by a description of one of those designs, the safe integral reactor (SIR). The rest of the article is then devoted to the international reactor innovative and secure (IRIS) which is currently being developed by an international consortium led by Westinghouse and including 20 + members from 10 nations. In discussing how IRIS has furthered the integral reactor concept, the design characteristics of IRIS are described along with its unique approach to safety, which focuses on exploiting the integral configuration to eliminate by design the occurrence of accidents in lieu of coping with their consequences. The IRIS safety approach is so successful that seven of eight class IV accidents (the accidents that could eventually lead to a radioactive release to the environment) typical of loop PWRs are either eliminated or downgraded, and thus it is being considered to pursue the licensing of IRIS without the requirement for off-site emergency response. The main characteristics of the IRIS safety by design are provided along with a discussion of its licensing approach.

Experimental study on an open cycle desiccant cooling system

This paper presents the design and construction of an experimental desiccant cooling system that uses natural zeolite as the desiccant. This is the first use of this desiccant in an actual or experimental system. The primary system components are originally designed and constructed. The paper provides a sufficient detail on the design procedure of the components as well as the methods of measurements. Experiments are conducted to cover a wide range of system and operating parameters. For the investigated cases, the coefficient of performance, the cooling capacity, and the moisture removal capacity take maximum values of 0.95, 16 kW/kg dry air, and 6.5 g/kg dry air, respectively. Exergy analysis is performed for a typical operation of the unit and the exergy destructions and exergy efficiencies are calculated. The greatest exergy destruction occurs in the desiccant wheel and the exergy efficiency of the unit is 3.3 percent.

DOE'S QUALITY SYSTEM PROGRAM: COOPERATIVE DEVELOPMENT AND IMPLEMENTATION

Implementation of a Quality Systems approach to making defensible environmental program decisions depends upon multiple, interrelated components. Often, these components are developed independently and implemented at various facility and program levels in an attempt to achieve consistency and cost savings. The U.S. Department of Energy, Office of Environmental Management (DOE-EM) focuses on three primary system components to achieve effective environmental data collection and use. (1) Quality System guidance, which establishes the management framework to plan, implement, and assess work performed; (2) A Standardized Statement of Work for analytical services, which defines data generation and reporting requirements consistent with user needs; and (3) A laboratory assessment program to evaluate adherence of work performed to defined needs, e.g., documentation and confidence. This paper describes how DOE-EM fulfills these requirements and realizes cost-savings through participation in interagency working groups and integration of system elements as they evolve.

Status and operational experience with acoustic monitoring of reactors primary systems

The paper presents diagnosis experiences by means of acoustic monitoring of German BWRs and PWRs. Substantial progress could be reached both in system technics and in application of acoustic monitoring with respect to the assessment of mechanical integrity of reactors primary system recently. During power operation of the plants acoustic signals of Loose Parts Monitoring System sensors are continuously monitored by dedicated digital systems for signal bursts associated with metallic impacts. ISTec/GRS's experience with its digital systems MEDEA and RAMSES have shown that acoustic signature analysis is very successful for detecting component failures at an early stage. Advanced powerful tools for classification and acoustic evaluation of burst signals have recently been realised. In a review some representative and characteristic cases of acoustic status evaluation of primary system components will be presented.

Nuclear district heating plants AST-500. Present status and prospects for future in Russia

Development activity in the field of nuclear reactors for district heating purposes was initiated in the ex-USSR in the middle 1970s. It was stimulated by a growing deficiency of fossil fuels needed to provide for the heating of large cities in the European regions of the country. Construction of two pilot twin-reactor nuclear district stations was started in the early 1980s in the cities of Gorky and Voronezh. Nuclear heating reactor AST-500, developed by OKB Mechanical Engineering, has a number of unique engineering features, e.g. integral lay-out of primary system's components, natural coolant circulation, steam-gas self-pressurizing system, additional safety vessel around the reactor unit, etc. All these specific features, which were adopted in order to assure enhanced safety of the reactor plant, required fulfilment of an extensive R&D program. Description of the reference AST-500 design features is given in this paper, as well as respective R&D activities carried out for the design solutions and safety validation. Currently, activity in the field is directed to AST-500 design upgrading and development and also to a whole series of heating-only and co-generation reactor plants ranging from 30 to 600 MW(th) for the new generation nuclear stations.

Differential labelling of primary olfactory system subcomponents by SBA (lectin) and NADPH-d histochemistry in the frog Pipa

SBA and NADPH-d histochemistries allow identification of functionally distinct components of the amphibian primary olfactory system. In Pipa, a secondarily aquatic frog, combination of both methodologies, using alternate sets of histological sections, reveals that, apart from Jacobson's organ, this species has a “water-nose” and an “air-nose”. The epithelia occupy separate chambers of the olfactory organ and give rise to olfactory nerve fiber bundles that are identified by the dual staining procedure.

US Nuclear Regulatory Commission research for primary system integrity regulations

Research at the Nuclear Regulatory Commission (Nrc) on the integrity and inspection of light water reactor (LWR) primary system materials and components, and on structural analysis, grew out of programs under way in the Atomic Energy Commission from the 1950s to the early 1970s. The research program was focused to provide a confirmatory and independent technical basis for rules, regulatory guidance and industry codes and standards that applied to the fabrication, inspection, analysis and safe operation of the materials and components of the primary systems in US LWRs. This paper provides an overview of the technical and regulatory issues dealt with over the 20 years of the Nrc's existence, with references to the detailed research which formed the technical underpinnings for the resulting regulations and regulatory guidance, and its impact on US codes and standards.

Primary coolant chemistry in VVER units

Operation experience with VVER coolant chemistry has been reviewed. The paper describes the results of measurements of radiation fields in primary system components and occupational doses that are compared with radiation control philosophy based on boron-potassium-ammonia coordinated water chemistry, modified and hydrazine water chemistry. The difference in water chemistry guidelines between VVERs and different current operational practices at VVER utilities are outlined. Special emphasis is given to the ammonia-hydrazine water chemistry on some VVER plants, high temperature filtration and silver behaviour.

Primary coolant chemistry in VVER units

Operation experience with VVER coolant chemistry has been reviewed. The paper describes the results of measurements of radiation fields in primary system components and occupational doses that are compared with radiation control philosophy based on boron-potassium-ammonia coordinated water chemistry, modified and hydrazine water chemistry. The difference in water chemistry guidelines between VVERs and different current operational practices at VVER utilities are outlined. Special emphasis is given to the ammonia-hydrazine water chemistry on some VVER plants, high temperature filtration and silver behaviour.

Acoustic monitoring systems—system concept and field experience

Abstract Continuous monitoring of primary system components, and acquisition and evaluation of factors detrimental to component parts, are important means to ensure safe operation and high availability of nuclear power plants. This report concentrates on diagnostic and monitoring systems based on acoustic methods. These systems are able to detect for instance loose parts, leaks and cracks. Systems for leakage and loose parts monitoring are described. Those systems have been developed, delivered and installed in numerous power plants all over the world. KWU is engaged in this area since more than 15 years. Crack monitoring systems based on the use of acoustic emission are still an object of research and development. Hence this technique is not part of the report. The paper describes in detail the PC-based System PC-KIDS for loose parts monitoring and the microprocessor-based system ALUS 86 for leakage monitoring. These monitoring systems, together with others (e.g. systems for fatigue and vibration monitoring) serve to • - provide the operating personnel with information on plant integrity • - permit early recognition of deviations from normal operation and anticipated operational occurrences to enable initiation of remedial actions • - avoid endangering personnel and prevent damage to the environment or equipment, • - maintain the plant operational beyond the specific life expectancy of 40 years.

Design procedure for the primary circuit of the Konvoi plants

The loading data for the design and fatigue analyses of the primary system components are derived from the plant operating transients as specified in the Operational Manual, the accident analyses and from the external event requirements. The methods used, applicable codes and standards, and QA procedures produce a design concept that contains significant safety margins. Measurements conducted in pre-Konvoi plants and during commissioning of the Konvoi plants deliver additional verifications of the analyses.

Acoustic monitoring systems — System concept and field experience

Continuous monitoring of primary system components, and acquisition and evaluation of factors detrimental to component parts, are important means to ensure safe operation and high availability of nuclear power plants. This report concentrates on diagnostic and monitoring systems based on acoustic methods. These systems are able to detect for instance loose parts, leaks and cracks. Systems for leakage and loose parts monitoring are described. Those systems have been developed, delivered and installed in numerous power plants all over the world. KWU is engaged in this area since more than 15 years. Crack monitoring systems based on the use of acoustic emission are still an object of research and development. Hence this technique is not part of the report. The paper describes in detail the PC-based System PC-KIDS for loose parts monitoring and the microprocessor-based system ALÜS 86 for leakage monitoring. These monitoring systems, together with others (e.g. systems for fatigue and vibration monitoring) serve to • - provide the operating personnel with information on plant integrity • - permit early recognition of deviations from normal operation and anticipated operational occurrences to enable initiation of remedial actions • - avoid endangering personnel and prevent damage to the environment or equipment, • - maintain the plant operational beyond the specific life expectancy of 40 years.

LOCUS operating system, a transparent system

The primary system components and architecture of the LOCUS operating system and the Transparent Computing Facility are identified. The current level of TCF functionality is described. References to the LOCUS OS are used in areas where the fundamental concepts are described. The initial two sections outline the vision of the LOCUS Operating System, and provide definitions of key terms. The third section identifies the functionality provided by IBM's Transparent Computing Facility. The fourth section describes several of the key algorithms and basic system structure, and the last two sections provide insight into future directions for the technology, and summarize the material presented.

Fort St. Vrain operations and future

June 1989 resulted in the highest electric generation month in the history of the Fort St. Vrain plant. This culminated an excellent operating period involving the first half of 1988. The plant was then shutdown in July on a scheduled outage to make repairs in the main steam drive section of all of the helium circulators. This outage was significantly extended due to problems with moisture ingress into the primary coolant felt to be primarily from the core support floor section of the liner cooling system. 1989 also marks the 10th anniversary of Fort St. Vrain as a commercial nuclear power plant. Electrical generation form this plant throughout this period has not been good, generally because of the prototypical nature of key plant components, primarily the helium circulators. Also, throughout this period there has been substantial increase in operations, maintenance, and fuel costs which could not be offset by the sale of electricity from the plant. As 1988 progressed, it became more and more evident that the financial burden of continuing to operate Fort St. Vrain could not be maintained by Public Service Company of Colorado (PSC). On December 5, 1988, the Board of Directors made a decision to permanently shutdown Fort St. Vrain on or before June 30, 1990. The lessons learned and operational experiences from Fort St. Vrain have been substantial for the development of the Gas-Cooled Reactor Program and in achieving a safe, commercially viable plant design for the MHTGR. Superb fuel performance has not only provided PSC with a radiologically clean plant, but has been the key to achieving significant safety attributes in the licensing of the MHTGR. Equipment performance for Fort St. Vrain on key components and systems such as the steam generators, purification system control rod drives, and many others has been factored into the design of the MHTGR. Additional benefits are anticipated as the plant completes its operating phase and goes through defueling and decommissioning. As primary system components become available and are physically evaluated, further valuable data will be secured for the benefit of the MHTGR.