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The pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR) of 220 MWe are made of Zr–2.5Nb alloy manufactured either from Double Melted (DM) or from Quadruple Melted (QM) ingots. These pressure tubes are manufactured by hot extrusion, two stages of cold pilgering with intermediate annealing and autoclaving. To achieve good in-reactor performance, it is required to have minimum variability in the mechanical properties of the pressure tube across its length and between tube to tube. In this work, tensile properties and fracture toughness parameters ( J max , d J /d a and CCL determined as per ASTM E1820-11 standard) of unirradiated Zr–2.5Nb alloy pressure tubes manufactured from DM and QM ingots using samples obtained from front and back end of the tubes is presented. The mechanical properties were evaluated in temperature range of 25–450 °C and compared with the corresponding data reported in literature for CANDU pressure tubes. [ABSTRACT FROM AUTHOR]

Spatial oscillations developed in a large nuclear reactor like an advanced heavy water reactor (AHWR) needs to be controlled properly for their safe operation. These spatial oscillations in the neutron flux distribution are developed due to xenon reactivity feedback. If these oscillations are not controlled properly, then a serious situation may occur in the reactor core causing “flux tilting.” Further, situations such as on-line refueling might cause transient variations in flux-shape from the nominal flux-shape. For analysis and control of spatial oscillations in AHWR, it is necessary to design a suitable control strategy, which will be able to stabilize these oscillations. In this paper, a fuzzy sliding mode controller (FSMC) is proposed for spatial control of AHWR. The FSMC approximates the discontinuous control of sliding mode control (SMC) to minimize the chattering and to improve the overall performance of the system. Through the dynamic simulations, it is observed that the designed FSMC is able to suppress spatial oscillations developed in AHWR and performance is found to be better as compared to recently proposed approaches in the literature. [ABSTRACT FROM AUTHOR]

A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made. [ABSTRACT FROM AUTHOR]

Electricity generated from nuclear power plants is generally associated with low emissions per kWh generated, an aspect that feeds into the wider debate surrounding nuclear power. This paper seeks to investigate how life-cycle emissions would be affected by including thorium in the nuclear fuel cycle, and in particular its inclusion in technologies that could prospectively operate open Th–U-based nuclear fuel cycles. Three potential Th–U-based systems operating with open nuclear fuel cycles are considered: AREVA’s European Pressurised Reactor; India’s Advanced Heavy Water Reactor; and General Atomics’ Gas-Turbine Modular Helium Reactor. These technologies are compared to a reference U-fuelled European Pressurised Reactor. A life-cycle analysis is performed that considers the construction, operation, and decommissioning of each of the reactor technologies and all of the other associated facilities in the open nuclear fuel cycle. This includes the development of life-cycle analysis models to describe the extraction of thorium from monazitic beach sands and for the production of heavy water. The results of the life-cycle impact analysis highlight that the reference U-fuelled system has the lowest overall emissions per kWh generated, predominantly due to having the second-lowest uranium ore requirement per kWh generated. The results highlight that the requirement for mined or recovered uranium (and thorium) ore is the greatest overall contributor to emissions, with the possible exception of nuclear energy systems that require heavy water. In terms of like-for-like comparison of mining and recovery techniques, thorium from monazitic beach sands has lower overall emissions than uranium that is either conventionally mined or recovered from in-situ leaching. Although monazitic beach sands (and equivalent placer deposits) only form 30% of the overall known thorium ore deposits, it is expected that such deposits would generally be utilised first if thorium becomes a viable nuclear fuel. Overall, for these four nuclear energy technologies, the range of CO 2 (eq) emissions per kWh generated (6.60–13.2 gCO 2 (eq)/kWh) appears to be low in comparison to the majority of electricity-generating technologies. [ABSTRACT FROM AUTHOR]

A multi-dimensional computational fluid dynamics (CFD) one-phase model to simulate the in-core coolant circuit of a pressurized heavy water reactor (PHWR) of a nuclear power plant (NPP) was performed. Three-dimensional (3D) detailed modeling of the upper and lower plenums, the downcomer and the hot and cold leg nozzles was combined with finite volume one-dimensional (1D) code for modeling the behavior of all the 451 coolant channels. Suitable functions for introducing the distributed (friction losses) and concentrated (spacer grids, inlet restrictors and outlet throttles) pressure losses were used to consider the local pressure variation along the coolant channels. The special power distribution at each coolant channel was also taken into account. Results were compared with those previously obtained with a 0/3D model getting more realistic temperature patterns at the upper plenum. Although the present model is restricted to one-phase phenomena, the prediction of the local pressure and temperature along the channels allows for a preliminary identification of the location of incipient boiling by comparing with the local saturation temperature. The present model represents an improvement with respect to the previous 0/3D model. It corresponds to the necessary step before achieving a 1/3D two-phase model with which the pressure drop and subcooled boiling along the coolant channels as well as the overall reactor pressure vessel (RPV) void fraction distribution can be evaluated more accurately. [ABSTRACT FROM AUTHOR]

Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce above-TS height and diameters substantially (vis-à-vis mid-1980s 4-loop, PFBR RP design with freeze-inflatable combination) which assures a significant step towards enhanced safety, economy and life (60 y at 85% capacity factor) of future FBRs, supported by stability maximisation, stress-minimisation and ease of critical component alignment. Closeness of inflatable seal operating requirements (25 kPa—120 °C—23 mGy/h) to TS barrier representative conditions (70 kPa—110 °C—23 mGy/h) are combined with key findings from R&D on elastomer, seal and coating of Indian FBRs to illustrate that critical elastomeric sealing of MOX fuelled FBRs, Pressurised Heavy Water Reactors and Advanced Heavy Water Reactor could be unified taking 50:50 blend formulation of Viton GBL 200S and 600S (developed for inflatable seal) as cornerstone and finite element based design as facilitator. It is further indicated that minimisation of synergistic ageing load on the fluoroelastomer inflatable-backup sealing combination by sodium dip seal could ensure 1 replacement during reactor life. [ABSTRACT FROM AUTHOR]

The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days. [ABSTRACT FROM AUTHOR]

Loss of coolant accident (LOCA) in nuclear reactor has a low probability to occur, but may have significant consequences due to decaying heat of the nuclear fuel rods. During LOCA in Indian Pressurized Heavy Water Reactor (IPHWR), there is a mismatch between heat generation and heat removed from the fuel channels, that results in structural failure of the fuel channels. These channels settle down at the bottom of the Calandria vessel, forming debris bed. Decay heat is absorbed by the moderator present in the Calandria vessel. The moderator level decreases due to continuous evaporation, which exposes fuel channels gradually. An experimental setup has been designed and fabricated to simulate the debris bed scenario. Experiments are conducted at different moderator levels, in the Calandria vessel and at different rated power. The temperature of the exposed rods gets stabilized with time due to steam cooling. It is observed that, the temperature is well below the melting point of the cladding, PT and CT material. [ABSTRACT FROM AUTHOR]

Understanding the dissolution behaviour of plutonium rich MOX fuel is one of the major challenges in fast reactor spent fuel reprocessing. As an initial step, kinetics for the dissolution of Indian PHWR natural UO 2 sintered pellets in nitric acid was studied experimentally. In this paper we have reported the effect of initial nitric acid concentration and temperature on dissolution rate of sintered UO 2 pellets. The shrinking core model was used to correlate the experimental results. The apparent activation energy estimated from the temperature effect of the chemical reaction was found to be about 90.5 kJ/mol. The estimated apparent activation energy indicates that the dissolution rate of UO 2 pellets was controlled by chemical reaction under the experimental conditions. [ABSTRACT FROM AUTHOR]

A CANDU lattice cell has been modeled using the Los Alamos National Laboratory's MCNP 6 code and Atomic Energy of Canada Limited's WIMS-AECL 3.1. Models for the CANDU 37-element fuel bundle have included a CANLUB coating, as a carrier for the neutron absorbers. The objective is to improve CANDU reactor operating margins by adding small amounts (~1 g) of neutron absorbers to each fuel element. For CANDU natural uranium fuel bundle design, the results indicate that (a) the fueling transient (due to the xenon-free effect) could be significantly reduced using gadolinium oxide (Gd203), with no significant impact on fuel burnup, and (b) the reactivity peak (due to plutonium production) could be reduced using europium oxide (EU2O3), with minimal impact on fuel burnup. An appropriate mixture of Gd2O3 and Eu2O3 that will improve operation and safety margins while having a minimal impact on fuel burnup is determined. Reactivity and power calculations for various mixtures of Gd2O3 and Eu2O3 are reported here. It is concluded that ~180 mg Gd2O3 and ~ 1000 mg Eu2O3 (~4.9 X 10-3 wt%per bundle) are sufficient to suppress the refueling transient and lower the axial plutonium peak, with a 0.27% burnup penalty (which is a small impact). Fuel safety and performance are always important topics for a nuclear utility. This approach of a relatively simple application of burnable poisons to existing CANDU natural uranium fuel design offers the benefits of improving fuel utilization and safety margins. [ABSTRACT FROM AUTHOR]

Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situ Pro perty M easurement S ystem (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated pressure tube which has been removed from Kakrapar Atomic Power Station-2 after 12.76 Full Power Years of operation for post irradiation examination studies at Bhabha Atomic Research Centre. This paper highlights the basic theory of measurement, results from the qualification trials on Zr–2.5Nb material having different mechanical properties and reactor operated irradiated pressure tube. [ABSTRACT FROM AUTHOR]

The characteristics, performance and general operation strategy of parabolic-trough direct steam generation (DSG) loop in recirculation mode have been numerically studied. The controllable separation steam quality ( x sep = 0.2 – 0.8 ) , which characterizes the two-phase flow, and the corrected direct normal irradiance ( DNI corr = 200 – 1000 W m −2 ), the major weather impact, were chosen as two independent parameters. The results show that for a given DNI corr , the onset of evaporation is advanced with increasing x sep . The convective heat loss is less or equal to the radiative part at the inlet, then surpasses and finally could reach up to 1.6 times of the radiative part at outlet. The energy efficiency decreases along the loop regardless of DNI corr or x sep while the exergy efficiency varies diversely according to DNI corr in the superheating stage. This diverse variation is due to the opposite trends of energy level and heat gain. When DNI corr < 400 W m −2 , the stratified flow shows an apparently increasing probability to emerge over the loop with decreasing x sep , where the thermal-stress risk may exist. To quantify the probability, a novel flow pattern indicator is proposed. In addition, a general DNI corr - x sep -based operation strategy of DSG loop concerning both performance and safety is given, of which the merit is verified through a case study in typical summer and winter days in North China. [ABSTRACT FROM AUTHOR]

In this work a calibrated Bonner Sphere Spectrometer (BSS), together with ISO shadow cones, was used to quantify the total and scattered components of bare and heavy water moderated 252 Cf neutron fields. All measurements were performed with a BSS that was calibrated at the National Physical Laboratory (NPL), Teddington, UK, which is a global primary standard laboratory and world-leading facility for neutron metrology and neutron instruments calibration. The fields were characterized for source-spectrometer distances of 80, 100, 150 and 200 cm; and at heights of 103 and 200 cm from the facility floor. As expected, the scattered contribution was greatest at the farthest distance from the source and closer to the floor. Hence, at a distance of 200 cm and a height of 103 cm, the scatter added to the direct field up to 162% of the total neutron fluence and up to 61% of the ambient dose equivalent, while at the same distance and height of 200 cm above the floor, these values were up to 146% and 52%, respectively. In the case of heavy water moderated 252 Cf neutron fields, a shadow cone subtraction technique could not be implemented, however Monte Carlo simulations were utilized in order to differentiate between the direct and scatter components of the neutron fields. In this case, at a source–detector distance of 200 cm and a height of 103 cm, the scatter added to the direct field up to 148% of the total neutron fluence and up to 45% of the ambient dose equivalent, while at the same distance and a height of 200 cm above the floor, these values were up to 134% and 42%, respectively. [ABSTRACT FROM AUTHOR]

Recent Fukushima nuclear accident has triggered further awareness amongst reactor designers regarding enhancing the safety measures in a nuclear reactor. It has become important to analyse the capability of decay heat removal in a reactor to avoid radioactivity releases to the environment. Such a study has been carried out for an Indian PHWR. In a hypothetical severe core damage accident in PHWR, multiple failure of the core cooling system may lead to collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a debris bed. Due to decay heat generation, the debris ultimately melts down forming a molten pool inside calandria vessel. Calandria vessel is surrounded by calandria vault water that acts as heat sink. In order to study the extent of heat transfer from molten pool to surrounding water under severe accident condition, an experiment was carried out wherein a simulant material was poured inside a simulated calandria vessel immersed in the simulated calandria vault water. The amount of melt and water present in calandria vault scaled proportionately with regard to an Indian 700 MWe PHWR. Results show that as soon as the melt was poured in the vessel, a thick crust was formed on the inner calandria vessel, which reduced the heat transfer from the melt pool to vault water resulting high temperature gradient in melt. Even though the cylindrical vessel inner temperature was found to be very high, the water outside the vessel never boiled. When the cylindrical vessel was opened after the experiment, there was no gap observed between the vessel and crust. Numerical analysis was carried out to predict the temperature profile of the molten pool and the vessel which were in good agreement with experimental results. Results were compared for the crust growth rate and temperature profiles in the melt pool considering the decay heat and without decay heat. Results show that with no decay heat consideration, the crust thickness continuously increases with time and in case of decay heat generation, crust thickness is found to be a function of decay heat. The melt temperature is found to increase above a decay heat of 1%. [ABSTRACT FROM AUTHOR]

Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper. [ABSTRACT FROM AUTHOR]

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured. [ABSTRACT FROM AUTHOR]

The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW tank-type reactor with a compact core design. It is moderated and cooled by light-water and has a heavy-water reflector. The MITR is a unique test bed for a wide range of nuclear material and fuel research. In one of the recent fuel cycles, detectors found elevated neutron levels on the reactor top lid during the operation with the Water-loop facility installed. A preliminary investigation theorized that the neutrons were streaming out via a thin CO 2 gas gap, which is used as a thermal isolation space for the in-core experiment. The current study employs neutron transport simulations to investigate the physical interpretation of the observed phenomenon. The state-of-the-art general purpose Monte Carlo code MCNP5 is adopted as the computational tool. The calculation results identify the typical characteristics of the streaming effect by quantifying the fraction of upward-streaming neutrons. Efforts have also been made to explore feasible engineering options for mitigating this neutron streaming from the Water-loop facility. It is concluded that a borated guide tube is not sufficient to reduce the neutron level on the reactor top lid; however, a stepped thimble design is effective in reducing the neutron source contribution. [ABSTRACT FROM AUTHOR]

Xenon induced spatial oscillations developed in large nuclear reactors, like Advanced Heavy Water Reactor (AHWR) need to be controlled for safe operation. Otherwise, a serious situation may arise in which different regions of the core may undergo variations in neutron flux in opposite phase. If these oscillations are left uncontrolled, the power density and rate of change of power at some locations in the reactor core may exceed their respective thermal limits, resulting in fuel failure. In this paper, a state feedback based control strategy is investigated for spatial control of AHWR. The nonlinear model of AHWR including xenon and iodine dynamics is characterized by 90 states, 5 inputs and 18 outputs. The linear model of AHWR, obtained by linearizing the nonlinear equations is found to be highly ill-conditioned. This higher order model of AHWR is first decomposed into two comparatively lower order subsystems, namely, 73rd order ‘slow’ subsystem and 17th order ‘fast’ subsystem using two-stage decomposition. Composite control law is then derived from individual subsystem feedback controls and applied to the vectorized nonlinear model of AHWR. Through the dynamic simulations it is observed that the controller is able to suppress xenon induced spatial oscillations developed in AHWR and the overall performance is found to be satisfactory. [ABSTRACT FROM AUTHOR]

• Experimental measurement of coolant void reactivity. • Measurement of void reactivity in Thorium-Plutonium fuel based experimental cluster. • Measurement of void reactivity for various partial voiding conditions. • Comparison of measured void reactivity with theoretically estimated values from deterministic and Monte Carlo based reactor physics codes. An Advanced Heavy Water Reactor (AHWR) based on thorium fuel is being designed and developed in India. The design has many inherent safety features; the most important among them is its negative coolant void coefficient. An experimental facility, named AHWR Critical Facility, has been built to validate the physics design parameters. Coolant void worth which is an important safety parameter was measured using differential critical height in a thoria based experimental MOX fuel assembly placed in the center of the core. The voiding effects were experimentally determined with both water and high density polyethylene (HDP) coolant. Different fractional voids were obtained using combinations of HDP block of compatible shape and sizes. The aim of the experiment was to validate the code systems used in physics design and in particular, to assess their capability of modeling the voiding conditions in thorium based fuel cluster. The validation exercise was performed with both deterministic and stochastic methods. The theoretical estimates of the void worth were found in very good agreement with that of the measurement. The results have enhanced the confidence in the code system used for AHWR physics design. [ABSTRACT FROM AUTHOR]

• The PT/CT contact leads to the high heat transfer to moderator. • The highest temperature reached in the PT for 4500 W/m2 heat input in fuel bundle was 1898 °C. • For 4500 W/m2 heat input rate the volume fraction of moderator vapour was maximum. • Upto 4500 W/m2 moderator acted as an efficient heat sink. For an Indian pressurised heavy water reactor, a numerical investigation was performed to study the boiling of moderator in a fully voided single nuclear channel during Pressure tube- calandria tube contact condition by applying variable heat flux. The setup included calandria tube, moderator, pressure tube, fuel bundle and an enclosure. It was observed that because of the contact between calandria tube and pressure tube, a notable variation in the circumferential temperature distribution of calandria tube and pressure tube were observed pointing out that contact condition drastically affects the temperatures of top and bottom portion of the tubes. It was also observed that upon increasing the heat flux, the top to bottom temperature variation decreases for pressure tube but increases for calandria tube. The phenomenon of boiling of moderator started when 1650 W/m2 heat flux was given to the system as input and the rate of boiling increased as the heat flux was increased which resulted in increasing moderator vapour fraction in the enclosure and decrease in average moderator density. The film thickness of the moderator vapour on the outer circumference of the calandria tube varied with time, following a cycle of increasing and decreasing thickness. The result showed that upto 4950 W/m2 heat flux, the moderator acted as an efficient heat sink. If the heat flux is increased further, the moderator can fail to act as a good heat sink and eventually lead to fuel channel failure. [ABSTRACT FROM AUTHOR]

• The present study compares the thermal creep behaviour of different generations of the Zr-2.5Nb pressure tube used in Indian Pressurized Heavy Water Reactors. • Pressure tubes manufactured from new route were found to be more creep resistant than the tubes manufactured from old route. • Both grain size and aspect ratio of the new route tubes were higher than that of old route tubes. • Higher extrusion ratios incorporated in fabrication of new route tubes produced stronger texture, which resulted in better creep properties. • Stress exponent and activation energy values revealed that the creep deformation mechanism was dislocation creep mechanism (dislocation climb – rate controlling mechanism, dislocation glide – strain producing mechanism). Indian Pressurized Heavy Water Reactors (IPHWRs) uses Zr-2.5Nb alloy pressure tube (PT) in cold work and stress relieved condition. These PTs are subjected to extreme conditions of stress, temperature and irradiation. Since past three decades of PHWR operation in India, fabrication routes of PTs for IPHWR has undergone several modifications to enhance its in-reactor performance. The deformation of PT in PHWR occurs due to irradiation creep, irradiation growth and thermal creep. In the present study, the different generations of PTs used in IPHWR are tested for its thermal creep behaviour. The accelerated creep rupture tests were carried out using flat tensile specimens with its axis parallel to the longitudinal direction of the PT in the temperature range of 350 to 450°C under stresses in the range of 0.7 to 0.9 times of yield strength at the respective temperatures. The minimum creep rates were evaluated along with the stress exponents and activation energies. The values of stress exponent and activation energy revealed the underlying creep mechanisms responsible for the creep deformation. Microstructural examination was carried out to understand the need and efficacy of the new fabrication route. The observed thermal creep behaviour of all the generations of pressure tubes were rationalised in terms of their respective texture and microstructures. Indian Pressurized Heavy Water Reactors (IPHWRs) uses Zr-2.5Nb alloy pressure tube (PT) in cold work and stress relieved condition. These PTs are subjected to extreme conditions of stress, temperature and irradiation. Since past three decades of PHWR operation in India, fabrication routes of PTs for IPHWR has undergone several modifications to enhance its in-reactor performance. In the present study, the thermal creep behaviour of different generations of PTs used in IPHWR is examined by carrying out accelerated creep rupture tests in the temperature range of 350 to 450°C under stresses in the range of 0.7 to 0.9 times of yield strength at the respective temperatures. The study revealed that the rupture life of the PTs following the new routes were higher than that of old route at 350 and 400°C. At 450°C all PTs had similar rupture life. A similar correspondence was made with minimum creep rate. The tubes manufactured by the new routes were more creep resistant than the old route tubes. [Display omitted] [ABSTRACT FROM AUTHOR]

• A perturbation expansion for the low frequency space-dependent transfer function of a zero power heavy water moderated reactor core yields a good approximation for the transfer function at fractions of 1Hz or an initial guess for it above that. • Perturbation expansion explains experimental and computational results. • Limits of application of this perturbation are outlined. The space-dependent transfer function defines the response of the neutron flux in reactor core to a periodic perturbation at a point in the core. To compute it at low excitation frequency for the ZED-2 zero power reactor, we introduce a perturbation method. The ZED-2 reactor is used to investigate the properties of CANDU® lattices. In CANDU®, the low frequency noise (0.1–5 Hz) allows one to study mechanical oscillations of structures in the core, the reactor control systems, as well as delayed neutron effects. The method for computing the ZED-2 transfer function was inspired by the results of measurements in the ZED-2 reactor. They showed that, at low excitation frequency, the imaginary part of the space-dependent transfer function is, up to normalization by the steady-state neutron flux, the same over all the core, while the real part varies significantly with the position of the flux detectors or the perturber in the core. This method was then found useful to compute the response of the ZED-2 reactor core to low frequency perturbations where other methods were found hard to apply. This opens the path to generalization of this method to CANDU® reactors. A rough calculation shows for which types of thermal reactors this perturbation theory is not worth applying. Also, the perturbation theory introduced here can be applied to other problems described by linear operators having a second smallest eigenvalue larger in absolute value compared to the perturbation. A further development, using the perturber induced reactivity, was found to make the core response normalization unnecessary. The theoretical basis of that method is outlined. [ABSTRACT FROM AUTHOR]

• Design, operation, and history of the British Steam Generating Heavy Water Reactor (SGHWR), a tube-type boiling water reactor, revisited. • Neutronic burnup calculations in Serpent are coupled with thermal hydraulics in THERMO for a 3D pressure tube lattice. • Lattice physics and reactivity feedbacks studied both for boiling conditions in the tube and superheated steam cooled operation. • Void coefficient shown to be increasingly negative with burnup, assisted by Plutonium build-up. • Power coefficient of reactivity is negative both for nominal boiling conditions and superheated steam cooled operation. The Steam Generating Heavy Water Reactor (SGHWR) was of significant interest in the UK in the late 1960s and early 1970s but failed to be developed into a commercial programme. This paper reassesses the reactor physics of the SGHWR with modern day modelling tools: neutronics and thermal hydraulics are coupled using Serpent and THERMO. Such a coupling is required since the boiling coolant along the SGHWR pressure tube height will have a significant influence on the neutronics of the system. Results show that the coolant void coefficient is negative at nominal conditions. This is achieved by the important contribution of the coolant towards moderation. Plutonium build-up throughout burnup also has a beneficial effect on the void coefficient. Quasi-static analyses of mass flow changes, pressure changes, and overpower events are examined. Most importantly, the power coefficient is negative. Superheated steam coolant conditions are also simulated for a SGHWR pressure tube. Although the steam temperature coefficient is slightly positive, the power coefficient is also negative. [ABSTRACT FROM AUTHOR]

The effectiveness of a control algorithm in the operation of a Pressurized Heavy Water Reactor (PHWR) relies upon its ability to counter the step back condition accurately. This paper presents an enhanced crow search algorithm for designing a fractional-order proportional integral derivative (FOPID) controller implemented in power control loop of a PHWR under step-back condition. The proposed controller is applied on eight different interval fractional transfer function models of PHWR derived using Edge theorem as well as eight step back conditions of PHWR. Simulation results reveal excellent reference tracking capability of the proposed controller. The effectiveness of the proposed controller has been proved by performing a comparison among different error performance indices with the existing controllers in the literature. [ABSTRACT FROM AUTHOR]

The present paper aims at providing experimental validation for the use of the MORET 5 code for advanced concepts of reactor involving thorium and heavy water. It therefore constitutes an opportunity to test and improve the thermal-scattering data of heavy water and also to test the recent implementation of probability tables in the MORET 5 code. [ABSTRACT FROM AUTHOR]

An irradiation system, which works based on photodisintegration reaction of deuterium, has been evaluated for determination of heavy water in water samples. This irradiation system was designed by simulation with MCNPX code, considering some gamma point-like sources placed in the center of sample, and then counting the created neutrons by the long cylindrical BF counter. For increasing the neutron counting efficiency, an optimization was performed for thickness of various moderators considered between the sample and thermal detector. To investigate the practical effectiveness of the designed system, a model apparatus was built for routine use in measurement of a wide range of heavy water concentrations. It was successfully applied for weight percentages larger than 0.5 % to observe the acceptable detection sensitivity of DO with an error less than 0.24 %. [ABSTRACT FROM AUTHOR]

This article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically reconfigured. The analysis was performed for Iran's IR-40 reactor at Arak in support of negotiations with Iran, but the methods have application to future reactors that present similar nonproliferation challenges. Two methods are considered, and both retain identical power, thermal-hydraulic, and safety profiles as the original reactor design. The conversion options can be implemented at any time during the reactor's life. The two methods have competing effects on achievable burnup, and they can be combined to produce an optimized core that matches both the fresh-core reactivity and maximum burnup of the original reactor. For the IR-40 example, the optimized design produces weapon-grade plutonium at only about 19% of the rate of the unmodified reactor for the same power level. Additionally, a reactor so converted could not be readily converted back to natural-uranium fuel without replacement heavy water, and it would retain the ability to produce medical isotopes at rates that exceed the original design through the use of LEU targets. [ABSTRACT FROM AUTHOR]

The article reproduces a presentation on the role of senses in technology historiography originally delivered at the 2013 Annual Conference of the German Society of the History of Technology. Case studies are discussed including the history of the olfactory detection of hydrogen sulfide gas during the heavy water enrichment activities on the shores of Lake Huron in Gagetown and Iroquois, New Brunswick and the role of human awareness and sensory perception in historiography is described.

Acoustic emission technique (AET) has been used for on-line monitoring of pressure vessels with artificial defects in the laboratory and for monitoring structural integrity of storage vessels in heavy water plant. The laboratory studies are helpful for getting better insight into the deformation and fracture aspects in the vessels, towards better utilisation of AET for structural integrity of components. AE results generated during hydrotesting of the storage vessels were used to confirm the structural integrity of the vessels. AE was used to monitor compression loading of plain and fly ash concrete. This showed three stages due to crack closure/microcracking, steady crack propagation and unstable crack propagation. An attempt was made to compare the types of cracking in concrete using AE parameters. [ABSTRACT FROM AUTHOR]

Before commissioning, Pressurised Heavy Water Reactor (PHWR) containments are tested for their integrity at proof pressure and Loss-of-Coolant Accident (LOCA) pressure. During operation, as per the current practice for Indian PHWR containments, in-service Integrated Leakage Rate Test (ILRT) is done once in two years at about one third of LOCA based test pressure. Kaiga Generating Station unit-2 (KGS-2) has been in operation for more than 11 years. For the first time, a periodic ILRT at LOCA pressure (0.86 kg/cm 2 (g)) was conducted at KGS-2. During this ILRT, structural monitoring of pre-stressed concrete primary containment was carried out. The deflection and strains at critical locations in the IC were recorded during ILRT mainly to compare the same with the values obtained from analytical study and with those observed during the first ILRT. The strain and deflection measurements were taken using vibrating wire strain gauges (VWSG) and high precision dial gauges respectively connected to data-logger. This technical paper highlights the structural behaviour of KGS-2 primary containment after 11 years of operation. [ABSTRACT FROM AUTHOR]

In nuclear power plants, continuous efforts are made to ensure that plants are operated in safe, reliable and economical manner. While utmost care is taken during design, construction and commissioning of the structures, systems& components (SSCs), continued healthiness has to be ensured during operation phase in accordance with the design intent. This is primarily achieved through the establishment of a comprehensive life management programme of surveillance, condition monitoring, periodic In-service inspections (ISI) and maintenance, the purpose of which is to ensure that required safety margins are maintained for all important SSCs throughout plant service life. In any industry, the important mechanisms which lead to SSCs degradation are general corrosion, Flow accelerated corrosion, erosion, thermal effects, fatigue and mechanical wear & fretting. In a Nuclear Power Plant, in addition to these, degradation also occurs due to irradiation and creep. Therefore SSCs need to be designed with due consideration of all such degradation mechanisms. During operation phase, the limiting conditions for operation and requirements of surveillance & condition monitoring for all components important to safety are documented in Technical Specification. A plant specific ISI manual defines and elaborates the ISI internal inspection methods and acceptance criteria. The ISI requirements are decided taking into account best industry practices and operational experience and meet the requirements of existing codes and standards. Both these documents are approved by AERB. In pressurized heavy water reactors, major components covered in ISI programme are coolant channels, feeder pipes, steam generators, heavy water heat exchangers, pressure boundary components, relief valves etc. The inspection methods adopted include use of UT, ECT, DPT and visual inspection. In addition to functional checks for some of the components, ferrography, vibration monitoring and thermography etc are also utilized for rotating equipment. The SSCs where unacceptable indications are revealed by ISI are repaired, replaced or isolated. The base line data generated during Pre-Service inspection is used for comparing and trending of observations. The observations made during ISI are subjected to thorough review and analysis by qualified experts to obtain assurance that unacceptable degradation in component quality is not occurring and it remains fit for service. The containment system of a NPP plays a crucial role in minimizing dose to the public in case of an accident situation. The integrity of Primary and Secondary Containment is assessed by conducting integrated leak rate tests in every Biannual Shut Down. The concrete structure is also subjected to NDT checks on a specified frequency to ensure continued healthiness. Indian nuclear power plants have accumulated an operating experience of more than 380 reactor years of operation. Over the years, the programmes for surveillance, condition monitoring and in-service inspections have been improved significantly and match with best in the industry. As a result, there has been no major age related failure in any of the important SSCs and thus continued safe and reliable operation of NPPs is assured. This paper brings out the ISI & health monitoring methodology adopted at Indian NPPs, to ensure safe and reliable operation. [ABSTRACT FROM AUTHOR]

Among the 43 organisations which joined the SARNET2 FP7 project from 2009 to 2013, 31 have been involved in the activities on the ASTEC code. This paper presents a synthesis of the main achievements that have been obtained on the ASTEC V2 integral code, jointly developed by IRSN (France) and GRS (Germany), on development, validation vs. experimental data and applications at full scale conditions for both Gen.II and Gen.III plants. As to code development, while the current V2.0 series of ASTEC versions was continuously improved (elaboration and release by IRSN and GRS of three successive V2.0 revisions), IRSN and GRS have also intensively continued in parallel the elaboration of the second ASTEC V2 major version (version V2.1) to be delivered end of 2014. Regarding code validation vs. experiments, the partners have assessed the V2.0 version and subsequent revisions vs. more than 50 experiments; this extended assessment notably confirmed that most models are today close to the State of the Art, while it also corroborated the yet known key-topics on which modelling efforts should focus in priority. As to plant applications, the comparison of ASTEC results with other codes allows concluding on a globally good agreement for in-vessel and ex-vessel severe accident progression. As to ASTEC adaptations to BWR and PHWR, significant achievements have been obtained through the elaboration and integration in the future V2.1 version of dedicated core degradation models, notably to account for multi coolant flows. [ABSTRACT FROM AUTHOR]

The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases. [ABSTRACT FROM AUTHOR]

Neutron diffraction measurements were carried out at 25 °C for aqueous LiNO heavy water solutions, (*LiNO)(DO) where x = 0.1, 0.05 and 0.01, in which the Li/Li isotopic ratios were varied. Structural information on intermolecular nearest neighbor Li···DO interactions in the extensive concentration range was derived from the first-order difference function, ∆( Q), obtained from the difference in scattering cross sections between Li- and Li-enriched sample solutions. The nearest neighbor Li···O distance and coordination number for sample solution with x = 0.1 were determined to be r = 1.969 (8) Å and n = 4.12 (6), respectively, corresponding to the four-coordinated Li ion in the solution. On the other hand, those obtained for the solution with x = 0.01 are r = 2.00 (2) Å and n = 6.0 (2), respectively, indicating that hexaaqua Li is dominant in the dilute solution. These results clearly indicate that a concentration dependence of the hydration number of Li occurs in the aqueous solutions. [ABSTRACT FROM AUTHOR]

A Monte-Carlo parametric study was carried out to investigate the neutron absorption profile in a model of LR-0 reactor when it is moderated by different mixtures of heavy/light waters at molecular ratios ranging from 0% up to 100% D2O at increments of 10% in D2O. The tallies included; neutron absorption profiles in control rods and moderator, and neutron capture profile in 238U. The work focused on neutron absorption in control rods entailing; total mass of control rods needed to attain criticality, neutron absorption density and total neutron absorption in control rods at each of the studied mixed water moderators. The aim was to explore whether thermal neutron poisons are the most suitable poisons to be used in control rods of nuclear reactors moderated by mixed heavy/light water moderators. [ABSTRACT FROM AUTHOR]

A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor. [ABSTRACT FROM AUTHOR]

The article offers information on then nuclear reactor "Atucha II" that reached 50 percent of its power production to the Argentina System (SADI) in Argentina. It mentions that Atucha II, is based on natural uranium and heavy water. It mentions that the power plant is developed in under the construction licenses, standards and programs defined by the Argentina's Nuclear Regulatory Authority (ARN).

Highlights: [•] Highly non-linear model of AHWR is used for spatial power control. [•] A simple fuzzy-like PD (FZ-PD) control structure with robust rule base is developed. [•] Robust rule structure reduces the difficulties in design and tuning of controller. [•] Proposed FZ-PD structure shows robust and better transient performance. [•] Proposed FZ-PD controller is able to suppress spatial oscillations in AHWR. [Copyright &y& Elsevier]

Highlights: [•] Life limiting mechanism of low enriched uranium based advanced heavy water reactor. [•] The coolant channels need to be designed as easily replaceable for long life of the reactor. [•] To meet the above design feature rolled joint detachment system has been developed. [•] The system works on the principle of shock heating and fast cooling. [•] The paper describes the design basis and observations from the experimental trials. [Copyright &y& Elsevier]