Refurbishment and core conversion of the Apsara reactor

Apsara is a pool type reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than four decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments, etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its life and for overall upgradation of its safety features. During refurbishment it is also planned to incorporate the basic design features of the multi-purpose research reactor (MPRR) which is being developed at Bhabha Atomic Research Centre (BARC). This paper gives an account of design features and safety aspects of the existing and modified Apsara reactor.     

Multi purpose research reactor

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research & development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.     

Thermal–hydraulic studies related to the design of LBE neutron spallation target for accelerator driven systems

Recently, studies have been taken up in world's leading nuclear research institutes to develop accelerator driven systems (ADS). Our department has earlier proposed a one-way coupled fast-thermal reactor of 750 MW (thermal). This reactor requires current in the range of 1–2 mA for proton beam of 1 GeV. A suitable liquid metal lead bismuth-eutectic (LBE) target based on buoyancy as well as gas driven method has been designed for this reactor earlier. In this paper, detailed thermal analysis in the spallation and window region has been carried out to study the operability of the target from thermo-mechanical point of view. FLUKA and computational fluid dynamics (CFD) codes have been used for this analysis. The results indicate that, the temperatures, thermo-mechanical stresses are within the required values. The detailed analysis of this work is presented in this paper.     

Research and development studies on plant and core seismic behaviour for a fast reactor

This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA in co-operation with ANSALDO and ISMES for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary to satisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactors is also pointed out.     

Study of hydride blisters in Zr-alloy using neutron tomography

Formation of hydride blisters in Zircaloy pressure tubes of pressurized heavy water reactor (PHWR) is a major life limiting factor which hinders the safe and uninterrupted operation of the reactor. Nondestructive detection and evaluation of location and size of these blisters as well as hydride distribution in the matrix surrounding them may help in damage quantification and residual life extension. In this article we present the neutron tomography studies carried out on simulated hydride blister samples grown on Zircaloy tubes. Characterization on samples having various levels of hydrogen concentrations were also carried out for quantification of the detectability of our neutron tomography system. We could identify the spatial in-homogeneity of hydride concentration present in the samples. Quantitatively hydrogen concentration difference up to 25 wppm has been observed experimentally and calibrated against image intensity in the reconstructed image. This study establishes neutron tomography as a potential non-destructive evaluation tool for the estimation of the severity of damage in the integrity of the pressure tubes and provides valuable information about kinetics of blister formation.     

Prioritization of components important for safety and sensitive to ageing. Application for TRIGA reactor

Due to potential impact on the SSC performances, the identification of ageing effects and implementation of appropriate methods for mitigation of these effects represents an important preoccupation of many organizations and has received even more attention in the last years in the perspective of Long Term Operation. The efforts used for developing time-dependent reliability models for all SSC could be considerable and may outweigh the benefits, but they are not always necessary, mainly because in some cases, the maintenance, test, inspection and surveillance methods are good provision to mitigate the ageing effects. To efficiently use the limited resources, it is necessary to identify and prioritize the SSC that need time-dependent reliability models by using specific criteria. This paper contributes to the effort of performing an efficient process of evaluation of ageing effects using probabilistic safety assessment models. The paper presents the approach developed for SSC selection, and the results of its application carried out for two particular systems of the TRIGA research reactor.     

Design of a management information system for the Shielding Experimental Reactor ageing management

The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.     

Repatriation of Vinča RA reactor spent fuel

Abstract

The RA research reactor is located at the Vin?a Institute of Nuclear Sciences near Belgrade, Serbia. The reactor is a 6·5 MW, tank-type, heavy water moderated and cooled reactor of Russian design which commenced operation in 1959. After being temporarily shut down in 1984 for refurbishment, a final shutdown decision was made in 2002. Operations are underway to safely remove and repatriate the spent nuclear fuel (SNF) to the Russian Federation (RF), as well as to improve waste management throughout the Vin?a site and prepare a plan for reactor decommissioning. As a major activity within the Vin?a Institute Nuclear Decommissioning (VIND) Programme, the repatriation of over 8000 SNF elements containing 2·5 tons of uranium metal will significantly reduce nuclear proliferation and environmental safety risks confronting the current facility. Poor water quality in the SNF storage basins and degraded fuel integrity significantly challenge efforts to repackage and transport the SNF. This paper will focus on the activities related to SNF repackaging and shipment, report on progress, detail significant challenges and provide an overview of the fully integrated VIND project.     

Scientific paradigms of structural safety of aged plants—Lessons learned from Russian activities

The study of the effects behind the degradation of components and materials is becoming increasingly important for the safe operation of aged plants especially when it comes to life extension. Since the Russian nuclear community began to examine life extension issues nearly 15 years ago, there is much to learn from these pioneering studies. At the Ninth International Conference entitled ‘Material Issues in Design, Manufacturing and Operation of Nuclear Power Plants Equipment’ held in St. Petersburg, 2006, recent data were introduced regarding the ageing effects of mechanical properties of various kinds of steel and welding joints of Russian NPP components. The meeting was organized by the Central Research Institute of Structural Materials (CRISM) “Prometey” in cooperation with the IAEA and JRC-EU.In reviewing the recent data presented at the Ninth Conference, the authors believe that the paradigms of structural integrity issues in aged plants are now reasonably well established in (1) fracture mechanics and irradiation hardening of reactor vessels and core internals and (2) thermal ageing and annealing effects. However, the first author, G. Saji, believes that the current approach of low-cycle fatigue is still unable to prevent and predict environmentally assisted cracks such as demonstrated in the IGSCC issues in the down-comer pipes of RBMK plants and various steam generator corrosion issues. This fundamental flaw stems from design codes, which do not incorporate the basic knowledge of electrochemical corrosion mechanisms as represented by the corrosion current.     

Thermal hydraulic phenomena related to small break LOCAs in AP1000

Since the TMI accident in 1979, a lot of attention in the nuclear engineering field has been drawn to the small break LOCA issue, around which plenty of work has been done both experimentally and theoretically. Subsequent reactor designs have also been greatly influenced.As a Generation III + reactor that received Final Design Approval by U.S. NRC, AP1000 employs a series of passive safety systems to improve its safety. However, the thermal hydraulic phenomena related to small break LOCAs in AP1000 have not been fully understood and further studies are still required.This paper investigated the available literature and information on thermal hydraulic phenomena that occur during small break LOCAs in AP1000, which included the critical flow, natural circulation, counter-current flow limiting, entrainment, reactor vessel level swell, direct contact condensation and thermal stratification. In particular, the physical phenomena, theoretical and experimental research conducted in the past few decades, and prediction models as well as their comparison and evaluation for the thermal hydraulic phenomena related to the small break LOCAs in AP1000 were concluded.     

30 Years of experience in operating the BN-600 sodium-cooled fast reactor

Experience in operating the BN-600 sodium-cooled fast reactor during its nominal service life as well as its service life extension period, an additional 15 years, is described. Information is presented on the performance indicators which were achieved and deviations from the normal operating regime which occurred when the reactor was first started up. The degree to which they affect the safety and technical-economic performance of the facility is evaluated. It is concluded on the basis of an analysis of the BN-600 operating experience that sodium-cooled fast reactors have now been mastered commercially and that their prospects for further development are good.     

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Effects of solute elements on hardening and microstructural evolution in neutron-irradiated and thermally-aged reactor pressure vessel model alloys

Nanometer-sized Cu-enriched solute clusters containing Mn, Ni, and Si atoms are considered as the primary embrittling feature in reactor pressure vessel steels. In order to understand the effects of solute atoms Mn, Ni, and Si on hardening and cluster formation, reactor pressure vessel model alloys FeCu, FeCuSi, FeCuNi, and FeCuNiMn were irradiated at 290 °C in a research reactor. Thermal ageing at 450 °C was also carried out to compare with the results in the neutron irradiation. The addition of Mn resulted in larger hardening and higher cluster number density in both thermal ageing and neutron irradiation. In FeCu0.8NiMn alloy, the size distribution of Cu-enriched clusters formed in 62-h thermal ageing (almost peak hardening) was very similar to that formed in the neutron irradiation, indicating they are on a similar growing stage. But the average Ni and Mn composition in clusters formed in neutron irradiation was higher. A good linear relationship between hardening and the square root of cluster volume fraction for both neutron irradiation and thermal ageing data was found.     

研究堆运行许可证有效期延续申请审查策略

生态环境部第8号令《核动力厂、研究堆和核燃料循环设施安全许可程序规定》对核动力厂、研究堆和核燃料循环设施运行许可证件延续事项作出了新的规定。为推动我国研究堆老化管理标准体系建立,分析了我国研究堆延寿审查策略发展历程,结合高通量工程试验堆等研究堆运行许可证有效期延续申请审查工作中的几个关键问题,提出了以定期安全审查为主、重点依据老化管理并兼顾技术规格书审查及差异性审查的审查策略,研究成果为我国研究堆老化管理法规标准的建立提供了实践经验及理论指导依据。      

The nuclear safety project at KfK 1972–1986 — Objectives and results

In 1972 the light water reactor safety activities conducted at the Karlsruhe Nuclear Research Center (KfK) were combined under the Nuclear Safety Project (PNS). Its primary objective was to assess in quantifiable terms the safety reserves which are provided in nuclear power plant design in a conservative approach. While in the initial phase R&D work conducted under the project was largely characterized by investigations of the design basis accidents, mainly the loss-of-coolant accident, emphasis in the past decade has been shifted more and more towards severe core and core meltdown accident analysis. The activities comprise both theoretical studies and experimental investigations, often performed in adequate, large-scale facilities. All activities have been an essential part of the reactor safety research program of the Federal Ministry for Research and Technology (BMFT) and have been coordinated with a number of other programs conducted in Germany and abroad. This paper gives a broad overview of PNS contributions to LWR safety research in the past 15 years and summarizes the results, comparing them with the general goals defined. In conclusion, the attempt is made to give an outlook on remaining activities in LWR safety research being carried out by KfK.     

Regulatory strategy for effective application of periodic safety review system

In less than 10 years, the first commercial pressurized water reactor (PWR) plant in Korea will reach its official design life. As part of safety activities, developed countries have already implemented periodic safety review (PSR) or equivalent programs to check and improve the safety of operating nuclear power plants (NPP) during their plant life. At the end of 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program and to apply it to Korean operating NPP. Since Kori Unit 1 started the review for the first tentative application of PSR as a model case in May 2000, it is now progressing well. Management of aging is one of the major factors to be considered in PSR and life extension of a nuclear plant. This paper is intended to introduce the regulatory aspect and strategy of Korean PSR. The background and scope of basic PSR guidelines are described, and a summary of technical criteria for aging management, which shows a regulatory direction for PSR, is also presented.     

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ_ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10³Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.     

Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.     

Determination of the residual stress tensor in textured zirconium alloy by neutron diffraction

Results of neutron diffraction studies of crystallographic texture and residual stress tensor components in cold-worked and annealed cylindrical components made from E-110 zirconium alloy are presented. Those components are used as plugs in the fuel elements of the VVER-type reactors; the resident residual stresses influence the durability and safety of the fuel elements. The experiments were carried out on the neutron diffractometers at Dubna (the IBR-2 pulsed reactor) and Berlin Helmholtz–Zentrum (the BER II research reactor). It is shown that the samples have fiber texture that is changed considerably with annealing. The type I residual stress tensors for both samples were calculated by the BulkPathGEO model. The cold worked component has 136–166 MPa tensile residual stress in the radial direction and zero stress along the axial direction. Residual stress values in the annealed component are close to zero.