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.

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 H 2 instead of D 2 ) 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.

CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.

Among the various parts in a pressurized heavy-water reactor (PHWR), pressure tubes are of tremendous importance. This is because they withstand extreme both pressure and temperature differences that exist between the Primary Heat Transport System (PHTS) and the moderator. The pressure tubes also contribute to prevention of fission product release from the PHTS to the PHWR plant (together with end fittings and nearby parts including plugs). When a PHWR is given a 1% derating, half is due to the aging of the pressure tubes. The main concern with pressure tubes is decrease of the safety margin. Most of the reduction comes from the effects caused by radial expansion and axial sagging, which are belong to four major phenomena including the thinning and the elongation. More specifically, the fuel-pin temperature distribution changes for the worse if deformation of the pressure tube occurs. Because there is extreme irradiation inside the core, the tube content is exposed to high temperature and high pressure. Thus, the shape of the pressure tube is deformed as times goes on. In this paper, using modeling of a deformed pressure tube in three-dimensional space, the effects on the fuel, coolant temperature, and coolant density, were studied quantitatively. This included a neutronics effect explored using coupled neutronics and thermal hydraulics (T/H) calculations. Among the results, only marginal changes of the neutronics effects were observed. The T/H results, which included temperature and density of the fuel and the coolant, were not critical. Through this study, we are now able to determine in new ways, conventional derating values from a pressure tube.

Flow accelerated corrosion (FAC) is the main reason for Feeder pipe wall thinning in CANDU pressured heavy water reactor (PHWR). Atomic Energy of Canada Limited (AECL) made design modification for Third Qinshan Nuclear Power Plant (TQNPP) to mitigate Feeder pipe FAC rate by increasing Cr content to more than 0.2 wt % and control primary heat transport system (PHTS) pH value between 10.2–10.4, more than 4 years laboratory verification test proved the modification was effective. More than 10 years in-service inspection performed by TQNPP proved that real Feeder pipes FAC rate is less than designed value; TQNPP Feeder pipe can meet the design life and even extend the service life.

While most of the severe accident related vulnerabilities arising from the inherent 40 odd year old PHWR design are common with single unit CANDU reactors and a number are also shared with LWR designs of that vintage, an evaluation of a station blackout accident at a multi-unit CANDU station reveals significant challenges to accident management options and potentially unacceptable off site radiological consequences. Opportunities for design improvements are abundant but unfortunately mostly ignored with both accident progression and consequence assessments by the utilities presented in a distorted positive light in defiance of engineered realities and public safety. Over-pressure protection systems in all relevant reactor systems (PHTS, Calandria, Shield Tank, and Containment) are inadequate for decay heat, let alone for other anticipated severe accident loads. Early passive heat removal by steam generators after a station blackout can be compromised by primary coolant removal into a large pressurizer located well below the pump bowl. There are no emergency means of high pressure water addition to the steam generators or the heat transport system which not only has an inadequate steam relief capacity for over pressure protection such that an early containment bypass by steam generator tube ruptures is a possibility, but also lacks a method of manual depressurization for early accident mitigation. In absence of a retaining LWR like pressure vessel, the reactor cores would release fission products without attenuation into the box like containments that are at 48% per day leak rate at design pressure very leaky and at less than 1 bar design pressure, structurally weakest of all operating reactor containments. The reactor buildings around each individual reactor unit are inverted cup like traps for combustible gases. A large number of safety significant components like the steam generators, pumps and the reactivity control devices are all outside the containment envelope. The production of combustible Deuterium gas from over ten km of carbon steel piping and over 50 tons of Zircaloy can be extremely high making the installed numbers and types of PARS not only inadequate but as early ignition sources also dangerous. Improvements after Fukushima are perfunctory and the analytical methods in support of severe accident management guidelines are outdated and incomplete. A lax and uninformed regulatory regime blindly supporting an intransigent industry resisting basic design enhancements has further exasperated, like it did in Japan, the severe accident related risk from continued operation of these reactors. These conclusions are based on thirty years of working on severe accident related issues at CANDU reactors, conducting extensive design reviews and developing computer codes and analytical methods for accident progression and consequence assessments. It is hoped that open discussions by professional engineers would foster change in name of public safety. It is also feared that nothing will change unless an accident occurs.

Currently, Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs) and Pressurized Heavy Water Reactors (PHWRs) have the lowest thermal efficiencies compared to those of other nuclear-power reactors and thermal power plants. Therefore, the objective of this paper is to propose modifications to a generic PHWR to yield an increase in overall plant thermal efficiency. The focus of this paper is primarily on the secondary side of a Nuclear Power Plant (NPP) and is directly dealing with wet-steam High-Pressure (HP) and Low-Pressure (LP) turbine stages and a Moisture Separator and Reheater (MSR). Modifications of the HP and LP turbine stages are based on utilizing moisture removal stages (having up to 60% removal efficiency), which reduce the moisture content as the steam passes through those turbine stages. Reduced energy losses and an increase in mechanical efficiency due to lesser moisture content results in an increase in thermal efficiency. Furthermore, implementing moisture-removal stages in the LP turbine gives the ability to eliminate the reheater in the MSR, thus resulting in an increase of thermal efficiency due to both, the higher mechanical efficiency of an LP turbine and the redirection of live-steam previously used by the MSR to a HP turbine. To be able to show an increase in thermal efficiency based on these modifications of a generic PHWR, the Pickering CANDU-6 nuclear-reactor parameters were used as a reference case in the software, called DE-TOP. The modifications suggested in this paper can be applied to any NPP that uses a Rankine steam-turbine cycle on the secondary side (PWR, PHWR and/or BWR) and recommended for implementation during planned replacement of LP- and/or HP-turbine rotors for new construction of PWRs and PHWRs.

Currently, i.e., in 2016, 436 1 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%). Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature. As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors. Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.

Operating CANDU PHWRs present significant challenges with respect to their ability to mitigate accidents that are beyond the envelope of design basis drafted over 40 years ago. Today, consideration of severe accidents is a public as well as a regulatory requirement whose implementation begs serious reconsideration in an international coordinated effort. The PHWR enhanced vulnerabilities to accidents such as a sustained loss of AC power, as in Fukushima, arise not only out of the inherent design features but also out of the institutional arrangements that surround their licensing. For one, the reactors will, in absence of a containing pressure vessel as in PWRs, put fission product activity directly into the containment, sport multiple potential containment bypass vulnerabilities and produce copious amounts of flammable gases due to presence of large amounts of Zircaloy in fuel channels and carbon steel in feeders. The relatively thin walled, stepped, welded Calandria vessel into which the disassembled core debris will rest has potential to mechanically fail early, causing explosive and energetic interactions of hot debris with enveloping water. This can catastrophically fail the reactor structures. For severe accidents the containments, well designed for design basis accidents, are either small and weak as in single unit plants or unable to practically take any significant over pressure in negative pressure multi-unit reactor buildings that depend upon a single vacuum building, too small for a multi-unit severe accident. The paper presents analytical arguments in support of these observations, lists conclusions from a series of design reviews and discusses development of ROSHNI, a new generation PHWR dedicated computer code package for simulating an unmitigated station blackout scenario. It does not directly address the institutional issues that handicap a potent reduction of the residual risk posed by continued operation of these reactors without serious design upgrades but discusses the regulatory failures in this regard. It introduces ROSHNI, a newly developed severe accident simulation package that models the reactor core in a greatly enhanced detail necessitated by the variability amongst reactor fuel channels. For a single unit CANDU 6 reactor, the code simulates thermal-mechanical degradation of 4,560 fuel bundles in 380 diverse fuel channels individually (for a total of over 70,000 dissimilar fuel entities) and computes source terms into containment of flammable deuterium gas and fission products. A number of questions are raised about differences between Hydrogen source terms and mitigation measures that are being implemented for light water reactors and Deuterium specific reaction kinetics in generation and mitigation that must be clearly differentiated but ignored so far by PHWR operators. A discussion of effectiveness of certain severe accident specific design upgrade measures that have been implemented at some operating plants is also addressed. For example, potential for a smaller than optimum number (for severe accidents) of PARS units to actually cause Deuterium/Hydrogen explosions as an unintended consequence is discussed. Continued reluctance of CANDU utilities to address a long standing issue of inadequacy of reactor overpressure protection is also detailed.

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models. A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states. Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up. The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.

Indian nuclear power programme is being implemented in three stages taking in to consideration limited uranium resources and vast thorium resources in the country. The first stage consists of investing natural uranium in Pressurized Heavy Water Reactors (PHWR). This stage has the potential of 10 GWe. The second stage involves large scale deployment of Fast Breeder Reactors (FBR) with co-located fuel cycle facilities to utilize the plutonium and depleted uranium extracted from the PHWR spent fuel. This stage has a potential of about 300 GWe. In the third stage, effective utilization of the vast thorium resources is planned. Indira Gandhi Centre for Atomic Research (IGCAR) instituted in 1971 at Kalpakkam, is involved in the mission of developing the technology of FBR. A host of multidisciplinary laboratories are established in the centre around the central facility of the 40 MWt Fast Breeder Test Reactor (FBTR). Presently, the construction of indigenously designed MOX fueled 500 MWe Prototype Fast Breeder Reactor (PFBR) that started in 2003 is in advanced stage and commissioning activities are underway. The design of PFBR incorporates several state-of-art features and is foreseen as an industrial scale techno-economic viability demonstrator for the FBR program. Beyond PFBR, the proposal is to build one twin unit having two reactors, with each of improved design compared to PFBR, to be commissioned by 2025. Subsequently, towards rapid realization of nuclear power, the department is planning a series of metal fueled FBRs starting with a 500 MWe Metal fuel Demonstration Fast Breeder Reactor (MDFR-500) to be followed by industrial scale 1000 MWe metal fueled reactors. The paper discusses in detail the above aspects and highlights the activities carried out towards designing MDFR.

The Qinshan Phase III (PHWR) Nuclear Power Plant has two 728 MWe CANDU6 reactors and it was found that the pressurizer steam bleed valves operated frequently during on-power refueling. The pressurizer steam bleed valves started to leak internally due to steam cutting, and the internal leakage further lead to frequent operation of degasser condenser spray valves. By investigating the response of CANDU 6 reactors during refueling operations, it was confirmed that the fueling machine flow injection caused the pressurizer level to rise during online refueling, compressing the pressurizer steam space, which is the root cause of PHT system over-pressurization. Based on the root cause analysis, a solution was proposed by adding a feed-forward control term to the PHT system inventory control logic, and the feasibility of this improvement was verified in unit online tests. After implementing the design modification, the system performances also proved the optimized logic change is effective. The research on this subject is not only beneficial to improving the safe and reliable operation of the PHT system in CANDU 6 reactors, but also provides a reference for improving the design of CANDU reactors of the next generation.

Nuclear steam turbines can be classified into two categories, one for BWR reactors where some countermeasures are taken for radiated steam and water, the other is for PWR reactors and PHWR (CANDU) reactors where steam and water are not radiated. As for Low Pressure section, there is some difference in LP rotor end structure, and LP last three stage blade components can be applied to all reactor types. The trend in nuclear power equipment is in a direction of larger capacity. In response to this trend, longer last stage blade is required if same number of casing is kept to make nuclear turbines reasonably compact. This paper addresses some of the key developments and new technologies to be employed focusing on longer Last Stage Blade (LSB) development with Continuous Cover Blades (CCB), and other enhancements in product reliability and performance.

Thermosyphoning capability, also known as the natural circulation is an important nuclear safety indicator in NPP. In the absence of any forced circulation of the NPP primary system, the cooling of the fuel to remove the decay heat can be provided by thermosyphoning to the steam generator (SG). Based on the theory of natural circulation and simulated test, this paper will further identify thermosyphoning capability of Qinshan III PHWR (CANDU6). Analysis shows that, the ultimate reactor heat sinks are SGs and related safety valves in case of loss main pumps. Thermosyphoning capability was confirmed as effective and acceptable. Furthermore, loss of Power accident happened during Qinshan III Unit#1 normal operating condition. We will investigate the impact of reactor core cooling and validate thermosyphoning as the capability of Qinshan III PHWR heat sink on accident. Finally, this paper will give some suggestions to NPP operator’s actions issues with reactor core temperature difference, reactor inlet headers sub-cooling margin, reactor outlet header temperatures, and others. Meanwhile, paper also discuss on the improvement of CANDU future design.

PHWR (also known as CANDU) primary heat transport (PHT) system is to make the pressurized heavy water through the reactor fuel channel, take away the heat produced by the fission of uranium fuel (i.e. fuel cooling), through the reactor feeder pipes, into the Steam Generator. PHWR reactor feeder flow value are a very important parameters, which are stable and meets the requirements of the design, plays an important role in the NPP safe operation. This paper summarize the Qinshan CANDU reactor feeder flow measurement in practice, based on the feeder flow measurement principle and measurement data are analyzed and studied, including feeder flow measurement stages, site preparation of material and equipment, data measurement methods, data analysis methods, dealing with flow measurement uncertainties, and some other key point need to be attention. The Qinshan CANDU reactor feeder flow measurement schedules create the same type of CANDU units commissioning records in the world, which is not only shorten the commissioning critical path time, but also laid the foundation for the safety operation. The reasons for the Qinshan success are also to analyze in this article. Finally, this paper based on the experience of Qinshan CANDU feeder flow measurement, put forward some suggestions on organization and implementation to minimize human error and increase its efficiency, and hope for the help on the similar work area in the future.

The most painful lessons that engineers will learn from reviews of the Fukushima disaster relates to the unacknowledged failure of regulators, designers and utilities in better retrofitting existing reactors in a timely manner to better withstand and mitigate known severe accident related challenges to reactor and containment integrity. Commercial nuclear power reactors have operated for over 50 years, and the first severe accident progression studies and understanding of related phenomena began to mature over 30 years ago. However, some very basic accident prevention, mitigation and management measures have not been required by the regulators and hence not initiated by the utilities. In public reviews that inevitably followed, severe accidents in power reactors at Three Mile Island, Chernobyl and Fukushima have been often presented as site specific aberrations in design, operations, safety culture and acts of God beyond mortal imagination. If a sincere soul searching, regulatory overhaul and actual, effective, timely and far reaching measures are not taken for operating reactors, the world is bound to witness recurring severe core damage scenarios followed by series of studies that will predictably conclude that such severe accidents can happen only in other jurisdictions and in other designs and that ‘our’ reactors are ‘safe’ and only minor enhancements are required to placate the anxious public or inquiring regulators. The paper examines the role played by regulators and other Canadian stakeholders in failure to ensure that the so-called residual risk from operating nuclear reactors is minimized in a timely manner. It presents specific examples of known deficiencies in PHWR designs that may degrade a sustained loss of power from a recoverable outcome into significantly more severe consequences. It discusses ideas for potential design retrofits and reiterates the need for more open, concerted and cooperative efforts internationally in accident progression and consequence analyses and supporting experiments. Taking station blackout as example of an instigator of a severe accident, the paper discusses certain elements of the current PHWR designs that accelerate the onset and progression of core damage and present substantially degraded opportunities for mitigation and control. It demonstrates how in many cases, a sustained loss of power in a PHWR may cause a containment bypass with early and unacceptable off-site consequences. Many issues discussed in this paper are common to other reactor designs and regulatory jurisdictions.

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void Coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D© system code. Within the framework of the third Agreement “NA-SA – University of Pisa” a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.

Atomic Energy of Canada Limited (AECL) has two CANDU ® reactor products matched to markets: the Enhanced CANDU 6™ (EC6™)[1,] a modern 700 MWe class HWR design, and the Advanced CANDU Reactor™ (ACR-1000™), a 1200 MWe class Gen III+ design. Both reactor types are designed to meet both market-, and customer-driven needs. Some of the new features incorporated into the EC6 reactor include increased power output, optimized maintenance outages, more automated testing and an Advanced Control Room. Lessons learned through feedback obtained from the operating plants have been incorporated into the design, and equipment obsolescence has been addressed. This paper presents basic EC6 design improvements; AECL works with its customers to assess their individual design requirements. Excellent neutron economy, on-power refueling, a simple fuel bundle, and the fundamental CANDU fuel channel design provide the EC6 reactor with unsurpassed flexibility in accommodating a wide range of advanced fuels and fuel cycles in addition to the standard natural uranium. These advanced fuels provide the promise of extending resources, reducing waste and enhancing proliferation resistance.

Qinshan Phase III is the first commercial pressurized heavy water reactor (PHWR) NPP in China, and it uses CANDU-6 design developed by AECL. Based on plant design and operation experience, the event tree analysis model has been developed for both small break LOCA (SB-LOCA) and large break LOCA (LB-LOCA), which is an important aspect of operational Probabilistic Safety Assessment (PSA). Both SB-LOCA and LB-LOCA event tree analysis have been performed for Qinshan Phase III CANDU-6 PHWR NPP (TQNPC). And the event sequence development and plant damage status (PDS) were provided in the analysis. It reflects actual plant configuration and response under a certain event, and various break type and locations were also considered in the event tree analysis, e.g. Pressure Tube Rupture, Pressure Tube and Calandria Tube Rupture, Feeder Breaks, Pressurizer Relief/Steam Bleed Valves Fail Open, Liquid Relief Valves Fail Open, etc.