Prototype fast breeder reactor Research Papers (original) (raw)

2025, Archives of Materials Science

This paper deals with processes occurring in creep-resisting steels during long-term degradation at high temperatures. The mechanical and creep behaviour of these materials is compared to their microstructure. The experimental methods, including mechanical testing, creep tests, and structure measurement -light metallography, SEM/TEM, electron and X-ray diffraction, were used.

2025

This study presents the development and analysis of a pneumatic steering mechanism for the T-55 tank, addressing challenges inherent in traditional manual systems. These systems require excessive physical effort due to high resistance in linkages, leading to operator fatigue and reduced maneuverability. The proposed mechanism integrates a pneumatic cylinder and "rocker arm to convert linear motion into precise rotational control, enhancing steering performance and driver comfort. The design leverages compressed air as a lightweight, safe, and responsive medium, ensuring adaptability to diverse operational conditions. Structural analysis via finite element methods (FEA) confirmed the mechanism's durability, with the rocker arm exhibiting a maximum von Mises stress of 46 MPa, well below the material's yield strength of 200 MPa. Fatigue analysis further demonstrated the mechanism's capacity to endure over one million load cycles, ensuring long-term reliability. Dynamic simulations using MSC.ADAMS validated the system's performance. The piston stroke, ranging from 0 to 150 mm, allowed precise control of steering linkages. Motion analysis confirmed a free travel distance of 132.5 mm, aligning with practical requirements for T-55 steering systems. The pneumatic system also reduced operator effort by over 50% compared to manual systems, significantly improving operational efficiency. Compared to traditional systems, the pneumatic mechanism enhances maneuverability, enabling smooth directional changes in challenging terrains while reducing driver strain. Its modular design facilitates seamless integration with existing tank frameworks, minimizing modifications. This work demonstrates the potential of pneumatic systems to modernize tracked vehicle steering mechanisms, providing enhanced agility, reliability, and safety. The findings ensure that tanks like the T-55 remain highly effective in modern combat scenarios.

2025, histoire-cnrs

La sécurité a, depuis longtemps, été au coeur des préoccupations des ingénieurs et scientifiques qui ont développé l'énergie nucléaire, ce qui n'empêchera pas des remises en cause profondes. Les brevets Joliot avant-guerre décrivent ainsi les principes de fonctionnement d'un réacteur nucléaire et s'interrogent déjà sur les moyens de prévention des accidents. L'expertise nucléaire, ses méthodes, ses concepts, ses structures, sont l'oeuvre en France du Commissariat à l'énergie atomique. Créé en octobre 1945, le CEA est l'acteur exclusif des recherches et du développement des applications nucléaires pendant plus de dix ans. Au cours de cette phase, un nombre réduit de spécialistes a la charge à la fois du développement et de la sécurité de leurs projets. Ce fonctionnement va être remis en cause au milieu des années cinquante. La mise en place de normes et règlements en matière de sécurité par les organismes internationaux (AIEA, ONU, Euratom) 2 , mais aussi les premiers accidents comme celui de Windscale en Grande-Bretagne font prendre conscience de la nécessité absolue de formaliser les questions de sécurité au CEA, tant dans les concepts que dans les structures 3 . Fin 1957, le responsable scientifique du CEA, le haut-commissaire Francis Perrin, initie une réflexion sur l'organisation de la sûreté nucléaire. Alimentée par les exemples américain, britannique et canadien, elle aboutit à la création, en janvier 1960, d'une Commission de sûreté des installations atomiques (CSIA), chargée d'examiner la sûreté des installations en cours et à venir du Commissariat. Toute construction, mise en fonctionnement ou modification des conditions de fonctionnement d'installations doit désormais être soumise à l'approbation de la CSIA. La CSIA se réunira tous les trois mois sous la présidence du haut-commissaire entre 1960 et 1970 4 . C'est une sorte de « tribunal » au sein duquel « siègent » les différents L'expertise de la sûreté nucléaire en France La revue pour l'histoire du CNRS, 16 | 2007

2025

Traditional safety performance requirements for nuclear reactors have been developed for critical reactors, whose kinetics characteristics differ significantly from sub-critical, accelerator-driven nuclear reactors. In a critical nuclear reactor, relatively small amounts of reactivity (negative or positive) can produce large changes in the fission rate. In sub-critical reactors, the self-multiplication (k) decreases as the sub-criticality (1-k) increases, and the responsiveness to small reactivity changes decreases. This makes sub-critical nuclear reactors less responsive to positive reactivity insertions than critical reactors. Also, larger negative reactivity insertions are needed in sub-critical reactors to shut down the fission chain if the neutron source remains. This paper presents the results from a computational analysis of the safety performance of sub-critical, accelerator-driven nuclear reactors. Coupled kinetics and thermal-hydraulics models are used to quantify the effectiveness of traditional protection and control system designs in sub-critical reactors. The analyses also quantify the role of inherent, passive reactivity feedback mechanisms in sub-critical reactors. Computational results are used to develop conclusions regarding the most favorable and effective means for reactor control and protection in sub-critical, accelerator-driven nuclear reactors.

2024, BARC NEWSLETTER

MINIATURE SPECIMEN TECHNIQUE AS AN NDT TOOL ... FOR ESTIMATION OF SERVICE LIFE OF OPERATING ... Kundan Kumar, K. Madhusoodanan and BB Rupani Reactor Engineering Division, Bhabha Atomic Research Centre ... This paper was selected as the... more

2024, DOAJ (DOAJ: Directory of Open Access Journals)

Analysis of intragranular carbide precipitate in the Heat Affected Zone (HAZ) of Martensitic Stainless Steel (MSS) weldment was carried out. Low carbon grade martensitic stainless steel weldment subjected to four point bend test in simulated sweet crude environment was analysed with Transmission Electron Microscope (TEM). The optical microscopy of the failed sample revealed the presence of intergranular cracks on both sides of the weldment in the HAZ. Electron transparent sample for TEM was prepared from the HAZ of the weldment using extraction replica technique. The examination of TEM specimen in imaging mode revealed the presence of precipitates on grain boundaries. The compositional analysis of the precipitates was carried out with Energy Dispersive X-ray (EDX). The result of EDX analysis showed the presence of chromium and molybdenum, this suggests, the precipitates were carbides of the form M 23 C 6. The study therefore upholds sensitisation as the mechanism behind the intergranular cracks observed in the HAZ of the MSS weldment.

2024, Journal of Materials Science Research

Low carbon, nitrogen alloyed 316L(N) SS is an high temperature structural material for Fast Breeder Test Reactor (FBTR) applications. Laser welding is a non contact, low heat input widely accepted welding process for welding a wide variety of materials due to its advantages like deep narrow welds, minimum distortion, narrow heat-affected zone, excellent metallurgical quality, ability to weld smaller size, thinner and thicker components and increased travel speeds compared to other welding processes. Creep rupture tests have been carried out on laser welded 316L(N) SS joints at 650ºC. The Rupture behavior of these joints has been investigated at stresses in the range 180-220MPa. In the present paper an attempt has been made to present the results on the creep failure mechanisms of creep-ruptured laser welded 316L(N)SS joints. It has been observed that creep fracture occurred in an inter granular fashion at all the stresses that have been tested and creep cavitation was the dominant mechanism in controlling the creep rupture behavior.

2024, Przegląd Spawalnictwa

2024, Nuclear Engineering and Design

• We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology.

2024, Nuclear Engineering and Design

Thermal hydraulic studies have been carried out to understand temperature dilution suffered by coretemperature monitoring system of a sodium cooled fast reactor. The three-dimensional computational model is validated against experimental results of a water model. Jet mixing phenomenon as predicted by different turbulence models is compared and RNG k-ε model is found to be better than other models. A comprehensive parametric study considering: (i) effects of construction/manufacturing tolerances on thermocouple positions with respect to subassembly positions, (ii) thermal/irradiation bowing of subassemblies, and (iii) changes in core power profile during reactor operation cycles has been carried out. The studies indicate the maximum possible dilution in fuel and blanket subassemblies to be 2.63 K and 46.84 K, respectively. Shifting of thermocouple positions radially outward by 20 mm with respect to subassembly centers leads to an overall improvement in accuracy of thermocouple readings. It is also seen that subassembly blockage that leads to 7% flow reduction in fuel subassembly and 12% flow reduction in blanket subassembly can be detected effectively by the core-temperature monitoring system.

2024, Materials at High Temperatures

Creep deformation, damage and rupture behaviour of 304HCu austenitic stainless steel have been studied at temperatures 923, 973 and 1023 K over the stress range 100-240 MPa. The material exhibited relatively short primary stage of creep deformation followed by secondary (steady-state creep) stage and relatively extensive tertiary stage of creep deformation. The transient creep deformation is analysed based on the Garofalo relationship, ε ¼ ε 0 þ ε T 1 À exp Àr 0 :t ð Þþ_ ε s :t: The variations of (i) rate of exhaustion of transient creep ðr 0 Þwithsteadystate creep rate _ ε s ð Þ and time to onset of secondary creep and (ii) initial creep rate with _ ε s were found to obey power-law relationship with powers close to unity, thereby facilitating the development of master transient creep curve. The variation of _ ε s with stress and temperature obeyed Dorn's equation modified with back stress concept. The time to onset of tertiary creep is found to be proportional to rupture life (t r) while the damage tolerance factor λ ¼ ε f _ εs:tr decreased with increase in t r. In view of the prolonged tertiary creep, associated with microstructural change and intergranular creep cavitation, Kachanov-Rabotnov model has been used to assess the creep damage behaviour of the steel. The damage assessment coupled with finite-element analysis closely predicted the creep deformation and rupture life of the steel.

2024

controller is the most widely used control strategy in the Industry. The popularity of PID controllers can be attributed to their robust performance in a wide range of operating conditions and partly to their functional simplicity. The process of setting of PID controller can be determined as an optimization task. Over the years, use of intelligent strategies for tuning of these controllers has been growing. Biologically inspired evolutionary strategies have gained importance over other strategies because of their consistent performance over wide range of process models and their flexibility. The level control systems on Deaerator, Feed Water Heaters, and Condenser Hot well are critical to the proper operation of the units in Nuclear Power plants. For Precise control of level, available tuning technologies based on conventional optimization methods are found to be inadequate as these conventional methods are having limitations. To overcome the limitations, alternate tuning technique...

2024, Nuclear Engineering and Design

h i g h l i g h t s Decay heat removal from degraded core of a typical SFR is highlighted. Influence of number of DHXs in operation on PAHR is analyzed. Investigations on structural integrity of the inner vessel and core catcher.... more

2024, Journal of Nuclear Science and Technology

The Hitachi Training Reactor (HTR) provided with pulse operation capability was utilized for an experimental study on the excursion characteristics of light water reactors. Measurements were performed on reactor power, reactor period , released energy, fuel temperature, fuel cladding strain, pressure in water channel and water activities. The reliability of the instruments and devices employed was confirmed to be satisfactory through pulse operation tests repeated 300 times, with inserted reactivities up to 1.50 $ , with reactor periods down to 15 msec, and hot spot fuel temperatures up to 1,300-C. A description is also given of the instrumentation for the measurements and reactor operation. I.

2024, Regulation & Governance

This paper explores the nature of expert knowledge-claims made about catastrophic reactor accidents and the processes through which they are produced. Using the contested approval of the AP1000 reactor by the US Nuclear Regulatory Commission (NRC) as a case study and drawing on insights from the Science and Technology Studies (STS) literature, it finds that the epistemological foundations of safety assessments are counterintuitively distinct from most engineering endeavors. As a result, it argues, those assessments (and thus their authority) are widely misconstrued by publics and policymakers. This misconstrual, it concludes, has far-reaching implications for nuclear policy, and it outlines how scholars, policymakers, and others might build on a revised understanding of expert reactor assessments to differently frame, and address, a range of questions pertaining to the risks and governance of atomic energy.

2024, Nuclear Technology

design, fabrication, and testing of 70-MW(thermal) prototype models of both the double-wall straighttube and the single-wall helical-coil concepts is currently under way.

2024

A shake table of 100 MT, largest in India was established at Indira Gandhi Center for Atomic Research, Kalpakam for conducting seismic qualification experiments of large size components of Fast Breeder Reactor. The table is placed on massive block foundation. The site is collocated with major safety structures. Attenuation relations were evaluated for rock blasting to avoid blasting related damage to the structures and personal. This paper presents the details of design and construction methodology adopted for this structure.

2024, Geotechnical and Geological Engineering

Foundation floor mapping is essential for very important structures to provide permanent data set for geological interpretations. Prototype fast breeder reactor (PFBR) is being constructing at Kalpakkam, India. Nuclear Island Connected Building (NICB) is the main nuclear safety related structure of (PFBR), and is resting on a raft of 101.5 m 9 93 m size at about 20 m below the existing ground level. Geological floor mapping was prepared after the excavation. 2 m 9 2 m grids were used for sampling study and mapping of the walls and floor. The floor region of this raft foundation was seen to contain certain weathered zone with fractures of thickness of in the order of 2 m. No evidences of faulting/shearing were observed on the surface of the floor area. The entire floor area was made up of compact and sound rock mass. However, there were some isolated patches of soft weathered/altered rock and small fractures in the rock. The structural features observed during the mapping exercise indicated the need for consolidation grouting so that the entire floor area function as single rock mass. This foundation geological map will assist in making better interpretation of post-construction foundation instrumentation data.

2023, European Journal of Science and Technology

Monte Carlo simulations provide accurate results for the neutronic response of the system under consideration if modeling is performed appropriately since it has great influence on the results. Sensitivity analysis of modeling approaches for geometry and fissile material composition distributions in the reactor core was performed by taking ITU TRIGA Mark II Research Reactor into consideration. The method of defining fuel element positions in the core by using circular or hexagonal lattice was considered as one case and three different methods of lumping material compositions in the fuel elements was considered as another case since these approaches are used by deterministic codes hence the accuracy of deterministic codes were also investigated. The validation study showed that both MCNP and Serpent Monte Carlo codes resulted in good agreement with the experimental data. It was observed that the handling of fuel composition in different ways did not influence the results significantl...

2023, Nuclear Engineering and Design

A follow-up calculation was made on the accident of Reactor No. 4 of the Chernobyl Nuclear Power Plant based on the literature and accident reports published by the USSR. The analysis code system used had models peculiar to a pressure tube type reactor, of which the accuracy had been verified by the experimental facilities at the O-arai Engineering Center and the tests made at the "Fugen" Nuclear Power Plant. The analysis data were prepared based on plant specifications and its operation history obtained from those published literature and accident reports. The analysis was composed of (1) a calculation of the nuclear and thermal-hydraulic characteristics, and the graphite heating and temperature distributions which were the basic data for the follow-up calculation of the accident, (2) an analysis of the plant behavior before the test started, using these basic characteristics, and (3) a follow-up calculation of the power increase which occurred after the test started. The analytical results were found to agree well with the data published by the USSR. It was confirmed from these analyses that the main factors causing the accident were the increased enthalpy at the core entrance caused by the test made at low power level and the increased void fraction due to reduced coolant flow rate, in addition to the nuclear characteristics and performance of the control system peculiar to the Chernobyl Nuclear Power Plant.

2023

Prototype fast Breeder Reactor (PFBR) is a 500MWe pool type Sodium cooled nuclear reactor. Grid plate (GP) is a box type structure with top and bottom plates of 50mm thickness interconnected by 1758 sleeves. GP Supports the core sub-assemblies (CSA) in their Predetermined zones and allow their loading /unloading ensuring required verticality in them for proper operation of Transfer arm and absorber rod drive mechanisms. It also serves as a plenum to distribute the Sodium flow among different Subassemblies (SA) and resting points of SA on sleeve have been hard faced to avoid self welding and supports Inner Vessel (IV).The manufacture of GP involves machining of 6.8m diameter, with drilling & location tolerance of ±0.1mm.The GP is supported on core support structure (CSS) at its outer periphery and intermediate support at number of locations are also provided to meet the slope and deflection requirements .The GP consists of 4 inlet nozzles, with a pair of nozzles Connected by pipes (I...

2023, Welding in the World

2023, Journal of Materials Science Research

2023, IFAC-PapersOnLine

This paper presents a new approach for acoustic detection of sodium boiling in a Liquid Metal Fast Breeder Reactor (LMFBR) based on Autoregressive (AR) models. The AR models are estimated on a sliding window and classified into boiling or non-boiling models by comparing the on-line estimated values of their components to the predictions of their components from the environment parameters using linear regression. In order to avoid false alarms the proposed approach takes into account operating mode information. Promising results are obtained on the background noise data collected from the French Phenix nuclear power plant provided by the French Commission of Atomic and Alternative Energies (CEA).

2023, HAL (Le Centre pour la Communication Scientifique Directe)

France has a historical tradition of codifying rules and regulations into an elaborate corpus of public law applied by a powerful administration. However the nuclear industry seems to have long been spared this tradition. This analysis of the development and operation of the French system for regulating nuclear risks between 1960 and 1985 brings to light a suppleness of the first rules, standards and orientations for risk-management. This French exception has two explanations: the structure of the network of the institutions involved in regulations; and the political, industrial and social context in which the "small world" of nuclear safety evolved. This analysis stimulates thought about how the French risk-regulation regime is evolving in the current context.

2023

2023, AIP Conference Proceedings

We report on our ongoing project to calculate the leading hadronic contribution to the anomalous magnetic moment of the muon a HLO µ using two dynamical flavours of non-perturbatively O(a) improved Wilson fermions. In this study, we changed the vacuum polarisation tensor to a combination of local and point-split currents which significantly reduces the numerical effort. Partially twisted boundary conditions allow us to improve the momentum resolution of the vacuum polarisation tensor and therefore the determination of the leading hadronic contribution to (g − 2) µ. We also extended the range of ensembles to include a pion mass below 200 MeV which allows us to check the non-trivial chiral behaviour of a HLO µ .

2023, Journal of Nuclear Science and Technology

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i. e., in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% proof stress.

2023, Nature

OECD report itemizes risks of renewable energy production Paris ENVIRONMENTALISTS have been campaigning for years to alert governments to the long-term risks of burning fos~il fuels for energy production, but the so-called 'clean' alternatives of wind, wave , solar power and geothermal power are not without serious hazards , according to a report from the Organisation for Economic Cooperation and Development (OECD). The report takes the "important positive aspects" of renewable energy use as given. But while the negative impacts of fossil fuels or nuclear energy may be longterm and planet-wide, those of renewabies tend to be more localized and shorter-term. In most cases, they arise because renewable sources convert low energy densities and thus require huge collector areas. Hydroelectric power , which accounts for 21 per cent of energy production in OECD countries (1978 figures), like wind power, is not directly polluting, but may render large areas unusable for agriculture or recreation and can destroy local ecosystems. Biomass, especially firewood, which accounts for about 20 per cent of energy produced in developing countries can, however , have serious environmental impacts. Apart from deforestation, with its risks of soil erosion, burning firewood can produce airborne , inhalable particles that may be carcinogenic, as well as greenhouse gases. Large-scale waste incineration, on the other hand, is only as clean as the material burned. If the waste stream contains plastics or chlorinated products, highly toxic and often halogen

2023, EPJ Nuclear Sciences & Technologies

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

2023, Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory

The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235 U, 238 U) and accumulation of long-lived fission products (85 Kr, 90 Sr, 137 Cs, 151 Sm) and actinoids (238,239,240,241,242 Pu, 236 U, 237 Np, 241 Am, 244 Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors.

2023, AU Journal of Technology

The effects of soaking and coldworks on sensitisation of austenitic stainless steels were studied. Annealed and samples coldworked to 20% and 40% were subjected to sensitisation heat treatment at 650C for 24 and 72 hours. Metallographic analysis carried out showed that sensitisation occurred in both annealed and coldworked samples which manifested as ditches. Soaking for 72 hours showed more ditches than 24 hours. Intergranular ditches were more prevalent in annealed and lower coldworked samples. In contrast, heavily coldworked samples were more susceptible to transgranular carbide precipitation evident by ditches on slip bands and recrystallised grains

2023

II. The inception of the nuclear power programme (from 1940) 9 2.1 Military ambitions 10 2.1.1 The early pioneers (early 1940s) 10 2.1.2 The Atomic Project (mid-1940s) 11 2.1.3 Windscale reactors (from late 1940s to 1950s) 16 2.2 The civil nuclear power programme 17 2.2.1 Motivations for the nuclear programme and the creation of the AEA (1950s) 17 2.2.2 British reactor technologies (1950s to 1980s) 18 2.2.3 Plan to privatise the electricity industry... 21 2.2.4 ...and the withdrawal of nuclear from privatisation 23 2.3 Summary of the evolution of the UK nuclear sector 27 III. The fast breeder reactor dream: expectations for the future 29 3.1 FBR as the long term goal 30 3.1.1 Dounreay Experimental Fast Reactor, DFR (1950s to mid-1960s) 31 3.1.2 Prototype Fast Reactor, PFR (mid-1960s to 1970s) 33 IV. The long decline 38 4.1 AEA split-up and the erosion of institutional support for fast breeders? (1970s) 38 4.2 Towards a commercial fast breeder reactor (CFR) and international collaboration (1970s) 39 4.3 RCEP's 6th report-the "Flowers Report" (1970s) 41 4.4 Reprocessing, proliferation concerns and the Windscale Inquiry (late 1970s) 43 4.4.1 The institution of public inquiry and the run-up to Windscale Inquiry 44 4.4.2 Reprocessing, FBRs and proliferation fears 46 4.4.3 The outcome of the Inquiry: landmark of participatory decision-making or a symbol of opposition against the 'nuclear complex'? 47 4.5 CFR-an experimental or commercial reactor? (late 1970s) 49 4.6 "Thatcher the scientist" takes office: an interlude of optimism in the fast reactor community (late 1970s) 50 4.7 International collaboration-preparations for the 'fallback option' begin (late 1970s to 1980s) 51 V. The final coup de grace: changing external environment, Thatcher's political ambitions and economic liberalism (1980s) 52 5.1 The (poor) economics of nuclear power 'revealed': calls for greater financial accountability (late 1970s to 1980s) 55 5.2 European fast breeder project-the last glimmer of hope (mid-1980s to 1990s) 57 5.3 Explaining Thatcher government's decisions on fast breeders-economics, ideology, and politics 59 3 5.4 'Pure' economics or political interests and strong personalities? Coal industry, miners' strikes and leaders of the UK nuclear establishment 60 5.5 Withdrawal of the UK from international fast breeder collaboration (early 1990s) 62 5.6 Shutdown and decommissioning at Dounreay (mid-1990s) 62 VI. Explaining the rise and fall of the fast breeder programme 63 6.1 Forecasts, predictions and expectations 64 6.2 Technological Arguments: problems inherent or external to the technology? 67 6.3 External factors 68 6.3.1 Uranium prices 68 6.3.2 Oil crisis 69 6.3.3 Risks, safety and security: radiation and proliferation 70 6.4 The role of policy actors and policy entrepreneurs: AEA and CEGB 71 6.4.1 Role of the AEA: from hegemony to decline and loss of mission 71 6.4.2 Institutional arrangements 74 6.4.3 AEA: united or internally divided? 75 6.4.4 The AEA and the CEGB 75 6.5 Policy and Politics 76 6.5.1 British industrial policy-or the absence of it? 76 6.5.2 Declining political support 77 6.5.3 Privatisation, liberalisation and the 'economic dogmatism' 78 6.6 NGOs and public opinion 80 VII. Conclusions: what and who killed the fast breeder reactors in Britain? 82 7.1 Economics and technology: autonomy or dependence? 82 7.2 Hype, disappointments and 'the reality' 83 7.3 External events 83 7.4 Social & safety concerns 84 7.5 Political power play and declining political support 85 References 86

2023

Prototype Fast Breeder Reactor (PFBR) is 500 MW e pool type Sodium cooled Reactor . Presently this reactor is at advanced stage of construction at Kalpakkam. In prototype Fast Breeder Reactor (PFBR), the Reactor Assembly (RA) houses the main primary Sodium circuit in a single vessel called Main Vessel (MV). The Main Vessel supports the core, vessel internals and contains the radioactive primary Sodium and argon cover gas under all conditions. It absorbs the mechanical energy released during the core disruptive accident (CDA). In order to have high integrity against leak of Sodium, no penetration is provided in the vessel and all the pipes are taken out through the top shield. SS 316LN has been chosen as the structural material for Main Vessel due to high temperature strength and good compatibility with Sodium. Main vessel is huge in size 12.9m diameter, 12.8 m height, thickness varying from 25 to 40mm with self weight of 134 T and contains approx. 1 50 T of Sodium. The Main Vessel m...

2023, Annals of Nuclear Energy

A key element in the safety of any Nuclear Research Reactor design is the evaluation of the reactor's ability to withstand events that could reasonably be postulated to occur and, if unmitigated, could lead to core damage or radionuclides releases to the atmosphere. A first step to ensuring that the reactor design is sufficiently robust to withstand accidents is to identify a comprehensive list of IEs that might lead to core damage or radionuclide releases. This work seeks to present as comprehensive as possible the results obtained from identifying possible important initiating events (IEs) applied in the development of PSA Level-1 study for a 10 MW Water-Water Research Reactor (VVR). The methodology involves the listing approach and the IE screening and grouping methodologies and the focus was on internal IEs due to random failures of components and human errors with full power operational conditions and the reactor core was the radioactivity source. The results provided a set of IEs that were as systematic and as representative as possible, providing confidence to the completeness of PSA study. This study is one of the first few to address comprehensive steps to identify important IEs used in Level-1 PSA study.

2023, Energy Procedia

Electromagnetic Pumps have been used for pumping liquid sodium in auxiliary circuits such as fill and drain and purification circuits of sodium cooled fast breeder reactors. Despite their low efficiency these pumps are used in fast reactors because of their high reliability and low maintenance due to absence of moving parts. Besides, EM Pumps can be used for pumping impure sodium. IGCAR has developed electromagnetic pumps of various capacities and successfully used them in experimental facilities. This paper deals with the design, development and performance testing of a large electromagnetic pump called Annular Linear Induction Pump (ALIP). This 170 m 3 /h capacity ALIP is for use in PFBR Secondary Sodium Fill and Drain Circuit (SSFDC) and was introduced in the sodium circuit of SGTF for testing its performance. In this type of pump, a linearly traveling magnetic field is generated by means of circular windings placed spatially apart in slots and excited by 3-phase supply. This traveling field induces circulating currents in liquid sodium which generates secondary magnetic field. Interaction of primary magnetic field and secondary magnetic field produces pumping force on liquid sodium. The pump duct is made of SS316L. Pump winding is made up of copper with class H insulation. The pump is designed for 360V and for a maximum sodium temperature of 450ºC. The pump is a reflux type of pump with both inlet and outlet on the same side. The pump was tested in the cold leg of Steam Generator Test Facility (SGTF) and its performance characteristics were obtained. The efficiency of the pump was also calculated and compared with the theoretical value. The successful testing and operation of the pump in SGTF has indicated sound design and indigenous manufacturing capability. This paper describes the design data of the pump and details of the testing carried out in IGCAR

2023, Annals of Nuclear Energy

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam-Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1-0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than $1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ÔCRAFTÕ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by b factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be 61.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

2023, Annals of Nuclear Energy

A comparative study has been made on the mechanical energy released in a core disruptive accident resulting from an unprotected loss of flow accident (LOFA) in a medium sized liquid metal fast breeder reactor with oxide, carbide and metal fuels. The study is conducted by ignoring the passive safety features incorporated in the design so that the accident scenario culminates in an energetic disassembly of the core with large energy release. The paper further provides the salient features of the analysis and presents the results with suitable physical explanations.

2023, Annals of Nuclear Energy

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 • 10 À8 /de for failure of shutdown function in case of global faults and 4.4 • 10 À8 /de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 • 10 À6 /ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is <1 • 10 À3 /ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability.

2023, ndt.net

Neutron radiography can be used very effectively to investigate performance of nuclear fuels. Some of the details that can be resolved in irradiated fuels by neutron radiography are massive hydriding locations in the zircaloy cladding and... more

2023, Biosciences Biotechnology Research Asia

Radiation has become inseparable part of the living environment contributing as cosmic rays, terrestrial radiation, fallout from earlier nuclear accidents and testing, the increased use of diagnostic radiology etc., Since early ages of time, the homosapiens have lived with natural radiation and his system has been adapted for the surrounding radiation and its effects. But still, radiation and its effects gain attraction in the minds of the general public because of the baseless panic registered through passage of time regarding the uncertainties associated with the consequences of radiation and the knowledge inadequacy on handling of the same. Operating nuclear power plants are preferably considered as the elemental portal to elucidate the effects of radiationas a fairly large quantity of radioactive materials are being used as fuel in the reactors. The main reason of the so called menace is because of the strong reason that radiation is nonsensory. Only instruments can detect the presence of radioactivity and measure the level of the radiation field. The dose reduction techniques in the occupational environment involve three major concepts: 1. Time 2. Distance and 3. Shielding. Apart from these control measures and though there are various other protective measures to protect from contamination by using protective clothing, respirators, etc., the dose reduction scheme has to be built in the design of the nuclear power reactor. The intensive practice and strict compliance to the radiation protection aims at bringing the exposure level As Low As Reasonably Achievable (ALARA). The paper brings out the features of ventilation methodology adopted in the prototype fast breeder reactor and the effective identification of radiological zoning to contain the contamination. The paper also tries to justify that the appropriate design strategyhelps in effective dose reduction in operating nuclear power plants.

2023, Aerosol and Air Quality Research

This paper describes an experimental methodology developed in an Aerosol Test Facility (ATF) for sampling and analysis of sodium aerosol (metal vapour) from the cover gas region of a fast rector, and details of the validation of the methodology in an experimental sodium loop. The methodology involves; (i) sampling of sodium aerosols by drawing them without exposure to the atmosphere, (ii) trapping them in paraffin oil medium, and (iii) analyzing the paraffin oil for the determination of mass concentration by the conductivity method and size distribution by using a Mastersizer. Validation of the methodology is carried out in a sodium loop called the SILVERINA facility. The aerosol size distribution is found to vary from 1 to 12 µm, with the Mass Median Diameter (MMD) around 4.0 µm (σ g = 1.5), and the mass concentration is found to be ~9.50 g/m 3. The experimental results agree with the values found in the literature.

2023, Science & Global Security

This article explores the safety capabilities of the 500 MWe Prototype Fast Breeder Reactor that is under construction in India, and which is to be the first of several similar reactors that are proposed to be built over the next few decades, to withstand severe accidents. Such accidents could potentially breach the reactor containment and disperse radioactivity to the environment. The potential for such accidents results from the reactor core not being in its most reactive configuration; further, when there is a loss of the coolant, the reactivity increases rather then decreasing as in the case of water-cooled reactors. The analysis demonstrates that the official safety assessments are based on assumptions about the course of accidents that are not justifiable empirically and the safety features incorporated in the current design are not adequate to deal with the range of accidents that are possible.

2023, Nuclear Applications and Technology

A new calibration has been developed for the electromagnetic flowmeters located on the 14-in. lines of the primary sodium coolant system of the Enrico Fermi Atomic Power Plant. This new calibration has incorporated the experimental hydraulic characteristics of the primary sodium coolant loop, the voltage outputs of the flowmeters on the 14-and 6-in. lines, and the original calibration of the flowmeters on the 6-in. lines. Utilizing this new calibration, the system pressure drop was found to vary as the 1.9 'th power of the flow rate of the primary system. This relationship is in good agreement with theory. A comparison of the measured sodium flow using the new calibration with the calculated value from heat balance measurements showed good agreement, with an average deviation of 1.7%. (The "as-read" values from the flowmeters based on the previous calibration, which were developed using pump hydraulic characteristics and the pump affinity laws, were too high by an average of 10%.) This new calibration is now being used in the Fermi plant.

2023, Nuclear Engineering and Design

h i g h l i g h t s • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively.

2023, Journal of Nuclear Science and Technology

Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled Fast Reactors (SFRs) were conducted focusing on postfailure fuel relocation and freezing behavior. These tests supplied complementary information to the existing CABRI tests with a single-pin geometry. Based on detailed data evaluation and theoretical interpretation for the three-pin cluster tests, it is concluded that axial fuel relocation and freezing are dominated by local fuel enthalpy, and the relation between penetration length and local fuel enthalpy observed in these CABRI tests is basically applicable to the large-bundle condition. It is also clarified that a fuel/steel mixture tends to create tight blockages near the axial ends of the relocating fuel. Part of the fission gas released from the heating-up and melting fuel is expected to be trapped within the bottled-up region between the upper and lower blockages and will keep this region pressurized for a relatively long period.

2023, Nuclear Technology

A new approach to achieving fast reactor safety goals is becoming apparent in the U.S. Fast Reactor Program. Whereas the "defense-in-depth"philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety rather than relying on addon active, engineered safety systems. Intrinsic reactivity feedback mechanisms, based on fundamental nuclear cross section and material motion changes with temperatures, combined with passive methods to assure removal of decay heat, are being quantified and included in analysis techniques to demonstrate the exceptional robustness of current advanced liquid-metal-cooled reactor designs in the United States.