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Predicting the power distribution within nuclear fuel is essential for predicting reactor fuel performance, since power distributions can impact pellet temperature distributions and fission product transport and migration. Analytical expressions for radial power distribution in fuel pellets were sought using lattice physics calculations to generate data and a machine learning technique to find representative expressions. Analytical approximations can be useful in nuclear fuel performance codes, such as element simulation and stresses (ELESTRES)/ element simulation code in a loss of coolant accident (ELOCA) for providing very rapid predictions of power distributions with reduced computational effort and memory requirements, relative to using an embedded or coupled neutron transport/burnup reactor physics code. Radial power distributions were calculated a priori using lattice physics codes to model mixed oxide (MOX) 37-element fuel bundles in pressure tube heavy water reactors. Such advanced fuels are of interest for future fuel cycles. Several datasets were generated with different amounts of PuO2 and variable neutron energy spectrum. Results of preliminary studies with the least absolute shrinkage and selection operator (LASSO) regression machine learning method have obtained analytical fitting functions with a mean maximum relative error (MRE) of 0.056 and a maximum MRE of 0.152 on the test set. However, using LASSO to estimate the coefficients of a physically motivated modified Bessel plus an exponential function, results in a lower MRE (mean MRE 0.041 and maximum MRE 0.11) on the same test set. Further potential improvements in both the curve fit and the machine learning methods are discussed.

As analysts still grapple with understanding core damage accident progression at Three Mile Island and Fukushima that caught the nuclear industry off-guard once too many times, one notices the very limited detail with which the large reactor cores of these subject reactors have been modelled in their severe accident simulation code packages. At the same time, modelling of CANDU severe accidents have largely borrowed from and suffered from the limitations of the same LWR codes (see IAEA TECDOC 1727) whose applications to PHWRs have poorly caught critical PHWR design specifics and vulnerabilities. As a result, accident management measures that have been instituted at CANDU PHWRs, while meeting the important industry objective of publically seeming to be doing something about lessons learnt from say Fukushima and showing that the reactor designs are oh so close to perfect and the off-site consequences of severe accidents happily benign. Integrated PHWR severe accident progression and consequence assessment code ROSHNI can make a significant contribution to actual, practical understanding of severe accident progression in CANDU PHWRs, improving significantly on the other PHWR specific computer codes developed three decades ago when modeling decisions were constrained by limited computing power and poor understanding of and interest in severe core damage accidents. These codes force gross simplifications in reactor core modelling and do not adequately represent all the right CANDU core details, materials, fluids, vessels or phenomena. But they produce results that are familiar and palatable. They do, however to their credit, also excel in their computational speed, largely because they model and compute so little and with such un-necessary simplifications. ROSHNI sheds most previous modelling simplifications and represents each of the 380 channels, 4560 bundle, 37 elements in four concentric ring, Zircaloy clad fuel geometry, materials and fluids more faithfully in a 2000 MW(Th) CANDU6 reactor. It can be used easily for other PHWRs with different number of fuel channels and bundles per each channel. Each of horizontal PHWR reactor channels with all their bundles, fuel rings, sheaths, appendages, end fittings and feeders are modelled and in detail that reflects large across core differences. While other codes model at best a few hundred core fuel entities, thermo-chemical transient behaviour of about 73,000 different fuel channel entities within the core is considered by ROSHNI simultaneously along with other 15,000 or so other flow path segments. At each location all known thermo-chemical and hydraulic phenomena are computed. With such detail, ROSHNI is able to provide information on their progressive and parallel thermo-chemical contribution to accident progression and a more realistic fission product release source term that would belie the miniscule one (100 TBq of Cs-137 or 0.15% of core inventory) used by EMOs now in Canada on recommendation of our national regulator CNSC. ROSHNI has an advanced, more CANDU specific consideration of each bundle transitioning to a solid debris behaviour in the Calandria vessel without reverting to a simplified molten corium formulation that happily ignores interaction of debris with vessel welds, further vessel failures and energetic interactions. The code is able to follow behaviour of each fuel bundle following its disassembly from the fuel channel and thus demonstrate that the gross assumption of a core collapse made in some analyses is wrong and misleading. It is able to thus demonstrate that PHWR core disassembly is not only gradual, it will be also be incomplete with a large number of low power, peripheral fuel channels never disassembling under most credible scenarios. The code is designed to grow into and use its voluminous results in a severe accident simulator for operator training. It’s phenomenological models are able to examine design inadequacies / issues that affect accident progression and several simple to implement design improvements that have a profound effect on results. For example, an early pressure boundary failure due to inadequacy of heat sinks in a station blackout scenario can be examined along with the effect of improved and adequate over pressure protection. A best effort code such as ROSHNI can be instrumental in identifying the risk reduction benefits of undertaking certain design, operational and accidental management improvements for PHWRs, with some of the multi-unit ones handicapped by poor pressurizer placement and leaky containments with vulnerable materials, poor overpressure protection, ad-hoc mitigation measures and limited instrumentation common to all CANDUs. Case in point is the PSA supported design and installed number of Hydrogen recombiners that are neither for the right gas (designed mysteriously for H2 instead of D2) or its potential release quantity (they are sparse and will cause explosions). The paper presents ROSHNI results of simulations of a postulated station blackout scenario and sheds a light on the challenges ahead in minimizing risk from operation of these otherwise unique power reactors.

Advanced nuclear reactors use large pool of water inventory with an immersed heat exchanger to remove decay heat especially in the case of Station Black Out (SBO). The isolation condenser (IC) immersed in Gravity Driven Water Pool (GDWP) of Advanced Heavy Water Reactor (AHWR) is an example of such systems. Heat rejected by the IC is absorbed in the pool. As a result, water density decreases and moves towards the free surface of pool causing layers of hot water piling up over colder one giving rise to stratified water inventory. consequently, the pool at the free surface starts boiling before the grace period (7 days). In the present paper, thermal stratification has been modeled in a power to volume scaled experimental setup. The study is focused on investigating the effect of heater orientation on suppression of thermal stratification in the pool for both the cases of with and without shrouds around heat exchanger.

In pressurized heavy water reactors (PHWRs), during an unmitigated severe accident, the absence of adequate cooling arising from multiple failures of the cooling system leads to the collapse of pressure tubes and calandria tubes, which may ultimately relocate to the lower portion of the calandria vessel (CV) forming a debris bed. Due to the continuous generation of decay heat in the debris, it will melt and form a molten pool at the bottom of the CV. The CV is surrounded by calandria vault water, which acts as a heat sink at this scenario. In-vessel corium retention (IVR) through the external reactor vessel cooling (ERVC) is conceived as an effective method for maintaining the integrity of a calandria vessel during a severe accident in a nuclear power plant. Under the IVR conditions, it is necessary to ensure that the imparted heat flux due to melt is less than the critical heat flux (CHF) at the bottom of the calandria vessel wall. To evaluate the thermal margin for IVR, experiments are performed in a prototypic curved section of calandria vessel (25o sector) of calandria vessel to determine the CHF, heat transfer coefficient, and its variation along with the curvature of calandria vessel. The effect of moderator drainpipe on CHF and the heat transfer coefficient has also been evaluated. It has been observed that the imparted heat flux is much less than the CHF at the bottom of the calandria vessel.

Due to emerging climate change concerns coupled with increased global energy demand, eventually the world needs to move to low-carbon-emission electricity-generating sources—nonrenewables, such as nuclear power, and renewables, as hydro, wind, geothermal, solar, and tidal. The only source is nuclear power that is reliable, concentrated, and can be of large installed capacity and operate with high capacity factors. This paper updates and presents the current status of nuclear-power deployment and small modular reactors (SMRs) development in the world. Unfortunately, within last 9 years, electricity generation with nuclear power has decreased from ∼14% before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011 to about ∼10%. Therefore, it is important to follow up with and understand the latest trends in nuclear power including SMRs development.

Retention of corium inside the calandria vessel (CV) by externally cooling it by calandria vault water is essential to mitigate severe accidents in pressurized heavy water reactor (PHWR). The thermal failure of CV can be prevented by effective decay heat removal on the outer surface of CV using vault water, which depends on the heat transfer behavior from the outer surface of CV to the vault water. Determination of limiting heat removal capability of vault subcooled water through outer surface of CV is very important. Since the CV has a very large diameter and length, the bottom most part of the CV almost behaves as a flat plate with downward natural convection boiling heat transfer. The natural convection heat transfer is lesser on the flat surface as compared to the curved surface of the CV. Thus, the critical heat flux (CHF) on the flat surface under downward boiling condition is the limiting CHF of the CV under external surface boiling scenario. In order to estimate CHF in this configuration with local boiling, experiments were carried out on a downward facing SS304 L flat plate simulating the conditions of CV of 700 MWel Indian PHWR. The pool boiling CHF obtained in this study is also compared with other earlier works.

In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom of the calandria vessel (CV) and form a terminal debris bed, which generates decay heat. With time, the moderator evaporates and the terminal debris bed ultimately melts and forms a molten pool of corium. If corium breaches the CV and enters the calandria vault, large amounts of hydrogen and other fission gases may be generated due to molten core concrete interaction, which may pressurize the containment leading to containment failure. In addition, the passive catalytic recombiner devices may be incapable of managing such large amounts of hydrogen. Hence, in-vessel retention of corium is the only option to the avert progression of the accident. The heat removal capability of the CV needs to be demonstrated in order to attain the goal of in-vessel retention, to contain the corium during severe accidents. A lot of numerical analysis of heat removal capability of the CV has been done. However, experimental demonstration of in-vessel retention has been rarely presented in the literature, especially for PHWRs. In this paper, in-vessel retention at prototypic temperatures has been presented. Experiments have been carried out in scaled CVs. Different corium simulants have been used at elevated temperatures and corium coolability has been demonstrated.

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

Advanced heavy water reactor (AHWR) is a natural circulation-based, light water-cooled, heavy water-moderated pressure tube type of nuclear reactor. In AHWR, the steam separation takes place in horizontal steam drums purely based on gravity separation principle. It is a known fact that efficiency of gravity separation is affected by the carryover phenomenon, i.e., conveyance of water droplets by the steam. To minimize the carryover, it is advised to reduce the superficial velocities of phases. Lowering the flow velocities also results in to lower pressure drop, which is very much desired. However, careful attention must be given to carryover phenomenon during design. An experimental test facility known as air–water loop (AWL) simulating the scaled down steam drum of AHWR with air–water mixture has been designed and experimental work performed on carryover phenomenon is presented here. Comparison of measured entrainment fraction with existing correlations and other visual observations are described. Numerical simulations with Euler–Lagrangian method have been carried out for which droplet size distribution measured experimentally is used as an input.

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

The pressurized heavy water reactor (PHWR) technology was conceived in Canada and has moved to several nations for commercial production of electricity. Currently, 49 power reactors operate with PHWR technology producing nearly 25 GWe. The technology is flexible for adopting different fuel cycle options which include natural uranium, different mixed oxide (MOX) fuel, and thorium. The technology has made substantial improvement in materials, construction, and safety since its inception. PHWRs have demonstrated excellent performance historically. Their safety statistics are excellent. Indian PHWRs also have shown economic competitiveness even in small sizes, thus providing an ideal design for new entrants. While the technology features of PHWRs are available even in textbooks, the objective of this paper is to highlight the historical development and salient features, and innovations for further improvement in operation, safety and economics. Thus, this paper shall serve as a curtain raiser for the special issue “Pressurized Heavy Water Reactors (PHWRs) Safety: Post Fukushima.”

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

In Korea, pressurized heavy water-cooled reactors (PHWR) account for 17% of operating units and have taken an important role in providing national energy supply. The recent biggest issue in domestic PHWR community was the continued operation of the Wolsong-1 CANada Deuterium Uranium (CANDU) plant, which has recently been approved to operate for 10 more years after a 30 year design life. In relation to this issue, various actions from both post-Fukushima lessons and Wolsong-1 (WS1) stress test results are being taken. In KAERI R&D, the following topics are studied to support the basis for these actions. First, PHWR severe accident issues such as (1) primary heat transport system (PHTS) overpressure protection capability, (2) containment overpressure protection capability, and (3) bypass source term are evaluated. Second, a computer tool (called MAAP–ISAAC) has been modified and updated to support analyzing Wolsong severe accident issues. Third, a decision supporting tool, called Severe Accident Management Expert (SAMEX)–CANDU, has been developed to aid emergency response experts under severe accident conditions.

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.