CANDU Stations Research Articles

Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations

Derivation of Uncertainty Distributions for Channel Flow Rate and Fuel Critical Heat Flux Predictions for Best-Estimate Plus Uncertainty Analysis of Slow Loss-of–Reactor Power Regulation Accidents in CANDU Stations

Sensitivity analysis of severe accident calculations for the estimation of CANDU-specific OILs

CANDU-specific operational intervention levels (OILs), were determined through extensive severe accident calculations with MAAP-CANDU, assessing a variety of postulated scenarios in single and multi-unit CANDU stations. OILs are simple, field-measurement criteria indicating the need for protective response actions (e.g., evacuation, relocation, food restrictions), and are derived from estimated radionuclide releases for various accident scenarios. Detailed OIL(t,mix) functions relate field measurements to generic dose limits, and are affected by the radionuclide release mix. Sensitivities to parameters in the severe accident calculations can affect radionuclide release mixtures, which may alter the OIL(t,mix) functions and could require adjustment of the OILs. The two categories of source term sensitivities are i) MAAP-CANDU analysis assumptions, and timing of releases (i.e. evolution of radionuclide releases over the accident). Despite some variation in the OIL(t,mix) curves from the sensitivities, the chosen OILs are still bounding and ensure reasonable response to accident conditions while maintaining public safety.

Thermodynamic predictions of CANDU fuel volatilization and fission product behaviour under severe accident conditions

A comprehensive thermodynamic investigation has been performed on irradiated uranium dioxide CANDU fuel under representative severe accident conditions. Two cases have been explored in this work: Case I) the fuel is in contact with varying combinations of hydrogen and steam (i.e., mainly reducing), and Case II) the fuel is in contact with varying combinations of air and steam (i.e., oxidizing). The system thermochemistry has been computed for a wide range of gas mixtures for both cases, fuel to gas ratios, temperatures, and hydrostatic pressures. The calculations predict fission gas speciation and fuel volatilization based on equilibrium thermodynamics, including the formation of several secondary phases, changes in UO2±x stoichiometry, fission product solubility in various phases, melting, and vaporization. The overall objective of this work was to inform source term analyses in the industry standard toolset SOURCE under severe accident conditions of a CANDU station.

Assessment of Concept Feasibility of Combined Electrolysis and Catalytic Exchange in Decontamination of Low Deuterium – Tritiated Heavy Water

A CANDU type fission reactor maintains large amounts of heavy water inventory with desired isotopic purity for use in the moderator and coolant circulation loops. But substantial amounts of water escape out of reactor system pressure boundaries and fills into the containment building volume due to normal, off-normal, maintenance and emergency activities in the power station. The heavy water escape poses a radiological hazard owing to the tritium activity and thus needs to be processed in heavy water inventory management units and the nuclear power station effluent treatment facility. These escapes are collected using vapour/liquid recovery systems of power station and processed either in vacuum distillation based upgrading columns or treated directly in liquid effluent treatment units and discharged into environment. Whenever the escapes recovered are small in quantities and low in deuterium isotopic purity, chemical exchange based Combined Electrolysis and Catalytic Exchange process is a more viable option for decontamination over conventional distillation. This paper assesses the concept feasibility of using CECE process as an effective volume reduction method for decontamination of low deuterium-tritiated heavy water. The study proposes a scheme for decontamination considering a base case analysis in a typical CANDU station. It is theoretically proved that in comparison to the dilution methods followed in liquid effluent treatment units of power station, CECE process not only results in the recovery of valuable heavy water, but also in the reduction of diluent requirements by several orders of magnitude. Moreover, similar low radioactivity water effluents are expected during the operation of upcoming fusion reactors like ITER, DEMO facilities, etc. CECE based decontamination system shall be considered as an efficient effluent processing technology for nuclear material accounting and recovery of fuel isotopes in ITER, DEMO facilities also.

RELAP5 simulation of CANDU Station Blackout accidents with/without water make-up to the steam generators

In the event of a complete Station Blackout (SBO) of a CANDU reactor, the current safety practice is to initiate depressurization early in the transient once a lack of power has been declared. This requires operator actions to manually open the main steam safety valves (MSSVs) on the secondary side initiating a crash-cool procedure. The depressurization of the secondary side allows make-up water to be supplied to the steam generator (SG) secondary-side thereby extending natural-circulation driven heat removal from the fuel in the primary heat transport system. Once depressurized there are several additional water sources available to replenish the secondary-side inventory, and in the event that emergency power cannot be restored emergency mitigating equipment (EME) are available to provide alternative water make-up. The objective of this paper is to examine the processes and phenomena involved during and after crash-cooling and compare these results to cases where operator actions are not credited. Simulations are performed until such time as the secondary-side inventory is stabilized from alternative water sources or until it is depleted.A detailed RELAP5 model of a 900MW CANDU plant has been created including the primary heat transport system (PHTS), the feed and bleed system, the steam generator secondary side, the moderator system, and the shield water cooling system. The 480 fuel channels were grouped into 20 channels by elevation and channel power. The models were benchmarked against the 1993 loss-of-flow event at Darlington NGS and agreed with the station data within the reported measurement uncertainty. Then the models were used to simulate the Station Blackout accidents with loss of class IV, class III, and emergency power supplies. Five different scenarios with/without crash-cool and with different water make-up options are modeled and key sensitivities determined. The results show that the depressurization of the secondary side may create a situation where continuous natural circulation breaks down and intermittent buoyancy induced flows (IBIF) takes place. The RELAP5 predicted IBIF phenomena are discussed, as well as the limitations of the current RELAP5 code. The main focus of this paper is on the early stage of the accidents, i.e. when adequate steam generator secondary side inventory exists and where damage to the main heat transport system can be precluded. The results demonstrate that EME actions to maintain SG inventory are effective and ensure fuel and fuel channel integrity.

ICONE23-1054 ROSHNI : A NEW INTEGRATED SEVERE ACCIDENT SIMULATIONS CODE FOR PHWR LEVEL 2 PSA APPLICATIONS AND SEVERE ACCIDENT SIMULATOR DEVELOPMENT

Severe accident progression in a PHWR is strongly influenced by the large variability that exists amongst various fuel channels and fuel entities within channels. A best effort prediction necessitates consideration of all risk sensitive phenomena and processes in a detail that reduces uncertainties in rates as well as magnitudes of source terms. A review of available computational codes points to a need of a much greater than hitherto undertaken, detailed modelling of the horizontal PHWR reactor channels, fuel bundles, end fittings and feeders with an advanced, more reactor geometry specific consideration of solid debris behaviour in the Calandria vessel. A new computer code ROSHNI improves upon existing computational aides firstly in detail in which the reactor is modelled and further in its fuller consideration of important PHWR related severe accident progression pathways and phenomena. The paper details the code modelling approach used in capturing the important system response parameters. It is expected that the source terms of Deuterium gas and fission products can now be evaluated with lesser uncertainty and with greater degree of control available to the analysts. Users are able to perform parametric and uncertainty analyses with greater ease. Initial modelling of progression of a severe core damage accident is being targeted for a CANDU 6 reactor. However, the code structure is generalized to extend the ability to other PHWR designs including multi-unit CANDU stations and Indian PHWRs. After Fukushima, serious concerns have been raised about severe accident mitigation capabilities of all operating nuclear power plants including CANDU reactors whose design concept presents specific vulnerabilities not common with other reactor types. While operating organizations and their supportive regulating bodies are cautious, to the point of lethargy, about design upgrades to operating reactors, only a best effort modelling of realistic accident progression pathways can empower them with the information required to justify the required risk reduction upgrades to reactors that were never designed with severe accidents within their design basis. Reduction of uncertainties by better modelling can also assist in the development of more effective severe accident management options and better operator training as the lessons learnt from Fukushima must be taken seriously. Otherwise decisions on hardware upgrades will remain handicapped by larger than necessary uncertainties inherent in the available but obsolete PHWR reactor severe accident modelling techniques. This paper concentrates on the rationale for development of a new code; its overall modelling approach and details of reactor core modelling. In summary, the code models transient behaviour of each and all fuel channels; models all individual fuel bundles with each represented by 16 fuel and Zircaloy rings; computes the response of the associated end fittings and carbon steel feeders; considers fuel heatup during staggered boiloff in channels followed by differentiated dry heatup of each fuel channel in steam and deuterium with consideration of moderator depletion. In all, fuel temperatures are evaluated at about 80,000 locations with separate consideration of thermal behaviour of 380 pairs of end fittings and the same number of feeders whose thermo-chemical response to severe accident conditions has a potential of generating additional combustible gas, not previously evaluated. It considers disassembly and thermo-chemical behaviour of individual fuel bundles to multiple fields of debris first suspended over underlying intact channels and gradually relocated to the Calandria vessel with consideration of impact of about 100 in-core devices. It also considers effect of failure of the Calandria vessel under thermomechanical loads and evaluates energy and fission product source terms for containment response. Deuterium gas source

Signal de-noising methods for fault diagnosis and troubleshooting at CANDU® stations

Over the past several years a number of domestic CANDU® stations have experienced issues with neutron detection systems that challenged safety and operation. Intelligent troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities, which can aid current stations and be used for the future generation of CANDU® designs. Fault modelling approach using Fault Semantic Network (FSN) with risk estimation is proposed for this purpose. One major challenge in troubleshooting is the determination of accurate data. It is typical to have missing, incomplete or corrupted data points in large process data sets from dynamically changing systems. Therefore, it is expected that quality of obtained data will have a direct impact on the system's ability to recognize developing trends in the process upset situations. In order to enable fault detection process, intelligent filtering techniques are required to de-noise process data and extract valuable signal features in the presence of background noise. In this study, the impact of applying an optimized and intelligent filtering of process signals prior to data analysis is discussed. This is particularly important for neutronic signals in order to increase signal-to-noise ratio (SNR) which suffers the most during start-ups and low power operation. This work is complimentary to the previously published studies on FSN-based fault modelling in CANDU stations. The main objective of this work is to explore the potential research methods using a specific case study and, based on the results and outcomes from this work, to note the possible future improvements and innovation areas.

Fitness for service assessment of degraded CANDU feeder piping—Canadian regulatory expectations

Allowance for the continued operation of feeder piping at some Canadian CANDU stations, which is experiencing active degradation mechanisms, has been based primarily on augmented inspection practices and conservative fitness for service assessments. The major degradation mechanisms identified to date are: pipe wall thinning due to Flow Accelerated Corrosion (FAC) and service induced cracking due to Intergranular Cracking due to Stress Corrosion Cracking (SCC) and potentially Low Temperature Creep Cracking (LTCC) mechanisms. Given that currently available industry codes and standards do not provide sufficient guidelines/criteria for assessing the degradation of feeder pipes, the Canadian Nuclear Safety Commission (CNSC) has asked the utilities to establish feeder pipe specific procedures to provide reasonable assurance that the risk associated with the feeder degradation is maintained at an acceptably low level. In response to this requirement, the Canadian CANDU industry has developed and continued to update feeder fitness for service guidelines to provide evaluation procedures and industry standard acceptance criteria for assessing the structural integrity of the feeder pipes. The scope and frequency of inspections are determined based on the results of the fitness for service assessments taking into account the relative susceptibility of feeder pipes to each specific degradation mechanism. While industry practices for the management of degraded feeder pipes have, in general, been complied with the regulatory expectations, outstanding issues still remain. Major regulatory concerns include uncertainties associated with limitations in both the inspection techniques and the mechanistic understanding of the degradation processes, which can impede inspection planning and fitness for service assessments.This paper presents the regulator's view of the current situation with respect to degradation of feeder piping, its implications for nuclear safety and the regulatory expectations on industry's management of the critical ageing phenomena.

05/01287 Reactor noise measurements in the safety and regulating systems of CANDU stations

05/01287 Reactor noise measurements in the safety and regulating systems of CANDU stations

04/01845 Reactor noise measurements in the safety and regulating systems of CANDU stations: Glökler, O. Progressin Nuclear Energy, 2003, 43, (1–4), 75–82

The codes WIMSD/4 and BORGES—part of the MTR-PC code package—have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products.The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm149 followed by Gd157 and Cd113. The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe135.

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.The specific applications include the following areas:⊎ Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.⊎ Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.⊎ Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

Retention and excretion of 95Zr-95Nb in humans.

This note describes the retention and excretion of 95Zr-95Nb in humans based on a recent CANDU experience and a literature survey of reported cases. Two data bases, QUEST and INIS were used for the survey. Three reported cases were discovered: two for occupational exposures and one for public exposure from nuclear weapons fallout. Human lung retention from these three cases, plus whole body retention and some limited fecal excretion data from a recently occurred exposure at a CANDU station, were reviewed and tested against predictions based on ICRP Publication 30 model. Based on the fits of this model to the reported data it seems that the three occupational exposures exhibit class Y behavior while the public exposure exhibits class W behavior. For only one case is the chemical compound known with certainty: ZrO2. Zirconium oxides are currently classified as class W in ICRP Publications 30 and 54. This work confirms a suggestion that oxides of zirconium be classified as class Y and should be taken into account by the ICRP in its future publications.