Indian PHWRs Research Articles

Validation of computer code ‘ATMIKA’ against RD-14M Small Break LOCA experiments

The simulation of Small Break Loss-of-Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility is performed under the auspices of International Atomic Energy Agency (IAEA) as an International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the in-house developed system thermal–hydraulic neutronic computer code ‘ATMIKA’, extensively used to analyze postulated events in Indian PHWRs. RD-14M is an 11MW, full-elevation-scaled extensively instrumented thermal hydraulic Canadian test facility, possessing most of the key components of a CANDU (CANada Deuterium Uranium) primary heat transport system (PHTS). The loop configuration is similar to figure-of-eight geometry of a typical CANDU circuit and it is intended to reproduce the important geometric features of a reactor PHTS and the appropriate operating conditions. ‘ATMIKA’ prediction and its comparison against SBLOCA experimental results are compared in this paper. A specific SBLOCA experiment ‘B9006’ is selected for the Computer code ‘ATMIKA’ predictions. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection (ECI) and represents most complete SBLOCA test conducted in RD-14M that includes all the phases of the transient (blow-down, high-pressure ECI, secondary pressure ramp (crash cool), refill, low pressure ECI, exponential pump ramp, and natural circulation). This simulation demonstrates that ‘ATMIKA’ is adequately capable of predicting the break discharge, PHTS depressurization, channel flow rate, channel voiding, fuel sheath temperatures and high pressure core injection flow through ECCS accumulator and initiation of low pressure ECI for test B9006. ‘ATMIKA’ predicted results are compared with experimental results and it is seen that predicted results for all phases of transient are in good agreement with experimental results.

Integrated modeling of postulated rod drop incident in a PHWR

Reactivity devices of standard 220 MWe Indian PHWRs include adjuster rods (AR) and regulating rods (RR) which are meant for maintenance of steady power level, adjustment of global power level and distribution and provision of additional reactivity whenever required. There is an upper and a lower absorbing element at each AR/RR location. In a postulated event of mechanical failure of lower absorbing element, it drops out of the core and inserts positive reactivity in the core. Analysis of such an incident is presented in this paper. Time required for the rod to fall through the surrounding medium is estimated by integrating the equation of motion of the rod. The reactivity insertion rate so obtained is used to simulate the associated power transient. It is observed that due to high-reactivity insertion rate, reactor period decreased rapidly, leading to actuation of reactor shutdown system based on low reactor period. This prevents any significant rise in the fuel and clad temperatures. It is found that peak fuel and clad temperatures attained during such a transient are well below the safety limits.

Evaluation of Fracture Resistance Behavior of Zircaloy Fuel Clad Tubes of Indian PHWRs Using Experiments on Ring Specimens and Continuum Damage Mechanics Models

Continuum damage mechanics models are very popular in prediction of crack growth and fracture resistance behavior of low-alloy ferritic and austenitic stainless steel components of nuclear reactors with various postulated flaws under different loading conditions. However, literature regarding the application of the above models for prediction of fracture behavior of Zirconium alloys, which are used for manufacture of fuel-clad and pressure tubes etc., are very limited. These models are very useful for designers and safety analysts as the parameters of the models are truly material properties and are transferable from the specimen to the component level. In this work, the nonlocal version of the Rousselier’s damage model was used to predict the fracture resistance behavior of double-edged-notched-tensile specimens made from Zircaloy-4 material. Initially, the micro-mechanical parameters were determined from the testing of ring-type specimens. Subsequently, these parameters were used in finite element analysis of the double-edged-notched-tensile specimen in order to predict the crack growth behavior and the crack path under applied displacement-controlled loading conditions. The fracture resistance behavior obtained in terms of J-R curve was also compared with the corresponding J-R curves of an axially-cracked pin-loading-tension specimen. The results were also compared with similar data from literature wherever possible. From the above results, it can be concluded that the nonlocal Rousselier’s damage model is a suitable tool for prediction of accurate fracture resistance behavior of various Zirconium alloy components in the nuclear reactors in order to ensure structural integrity of the above components in various postulated accidental scenarios.

Testing and assessment of fatigue life prediction models for Indian PHWRs piping material under multi-axial load cycling

In the present paper, experimental and analytical fatigue studies on primary piping material of Indian PHWRs, have been carried out under multi-axial loading with an aim to assess the adequacy of currently used fatigue life models based on ASME B&PV Sec-III procedure and plastic strain energy density. The experiments were conducted under axial, torsion and combined axial–torsion loading with different loading waveforms, phase angle and strain ratio (shear to axial). The tests under varying principal directions loading resulted in shorter fatigue lives. The orientations of crack initiation plane were measured and found dependent on the loading parameters which supports the idea of critical plane based fatigue models.

Determining the age and history of plutonium using isotope correlations and experimentally determined data on isotopic abundances of plutonium and 241Am

Linear correlations using a data set of nominally cooled samples of Pu formed in Indian PHWRs using experimentally determined 241Pu/239Pu versus 240Pu/239Pu and 241Am/239Pu versus 240Pu/239Pu isotope ratios have been developed which can be used for determining the age of Pu. By correlating both Pu and Am isotopic information, an understanding of how the material was processed, when chemical separations occurred to remove 241Am as well as the true age of the Pu in sample can be obtained. Two actual samples from a PHWR with unknown origin were analyzed as a part of case study for application of this new methodology.

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

Experience with Dilute Chemical Decontamination in Indian Pressurized Heavy Water Reactors

Abstract Dilute Chemical Decontamination (DCD) process has been used in several full system and components of nuclear coolant systems to effectively remove the radioactive contaminants that causes radiation field and consequent MANREM problem. The DCD process uses chemicals in very low concentrations (millimolar) and dissolves the oxide film along with the activity incorporated in the oxide film. In DCD process operated under the regenerative mode, the chemical formulation spent in the process of oxide dissolution is replenished by passing through cation exchange columns. Finally, after achieving sufficient decontamination of the system/component, the added decontamination chemicals along with the activities and metal ions released during the process are removed by mixed bed ion exchange columns and the system is restored to normal operating condition in a few days time. In PHWRs, the regenerative DCD process is applied for full primary coolant system decontamination. The chemicals are added directly to the heavy water coolant with the fuel in the core. In Indian PHWRs (MAPS#1&2, RAPS#1&2, NAPS#1&2 and KAPS#1), the process has been applied eleven times. A chemical formulation based on NTA, Citric acid and Ascorbic acid has been applied seven times with good results. Decontamination factors in the range 2-30 have been obtained in different components with good MANREM savings in the subsequent maintenance works. Efforts are on to modify the process to take care of the challenges posed by antimony isotopes. An inhibitor (Rodine-92B) based process was successfully tested in NAPS#2 for removing antimony isotopes (122Sb and 124Sb). Further refining of the antimony removal process is being worked out. Similarly, the process is being modified to effectively remove the hotspot causing stellite particles in the moderator system of PHWRs. A permanganate based process has been developed and tested in several adjustor rod drive mechanisms in KAPS and NAPS. The experience of applying/testing the DCD process in the Indian nuclear reactors is described in this paper.

A diagnostic system for identifying accident conditions in a nuclear reactor

The objective of this study is to develop a system, which assists the operator in identifying an accident quickly using ANNs that diagnoses the accidents based on reactor process parameters, and continuously displays the status of the nuclear reactor. A large database of transient data of reactor process parameters has been generated for reactor core, containment, environmental dispersion and radiological dose to train the ANNs. These data have been generated using various codes e.g., RELAP5—thermal-hydraulics code for the core. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs. The system has been designed to provide the necessary information to the operator to handle emergency situations when the reactor is operating. The diagnostic results obtained from ANNs study are satisfactory.

Development of polymeric applications for sodium cooled Fast Breeder Reactors: Chronicles of inception, progress and achievements

The collaborative programme on development of important polymeric applications of Indian FBRs is chronicled from the days of motivation to its present state. Failure of inflatable seals of FBTR RPs (1985) and adoption of all-elastomer sealing concept for PFBR RPs (early 1990s), coupled with the unique characteristics of elastomeric materials, led to inception of the programme at IGCAR (1998) which involved DMSRDE as the first partner (1999). The planned initiative, which eventually involved more than 15 other Indian agencies, resulted in complete development of FKM backup seals for PFBR RPs which has been installed in reactor recently. Coated FKM and EPDM inflatable seals for PFBR and FBTR RPs have been developed, produced and evaluated up to ∼2 m diameter. Development methodologies for other critical polymeric applications of PFBR, FBTR and FCF have been formulated. Accomplishments and novelties of the development include EPDM and FKM compounds and designs for inflatable and backup seals, a common FEA procedure for elastomeric ring seals, PECVD based Teflon-like coating technology up to 7 m seal diameter, seal production process by cold feed extrusion and continuous cure, a robust quality control framework and the new facilities developed to support the programme. Future developments are focused on delivery of validated inflatable seals, life assessment and development of new elastomeric compounds which include silicone rubber and perfluoroelastomer, PECVD based coating on stainless steel and development of adhesionless joining of FKM. The achievements and future research will standardize the design and development of the elastomeric seals of Indian FBRs, PHWRs and AHWR based on a few well-characterized compounds, a common FEA method and PECVD based coating technology which can result in a universal design code.

Fatigue studies on piping components including fatigue-ratchetting interaction

A component test programme was initiated to address the applicability of leak-before-break to Indian PHWRs. After completing the tests with constant amplitude loading under air environment, the interactive nature of fatigue was addressed. The parameters considered the present studies are: Water environment and constant amplitude loading Loading due to bending and torsion Fatigue ratchetting

Optimisation of the hot conditioning of carbon steel surfaces of primary heat transport system of Pressurized Heavy Water Reactors using electrochemical impedance spectroscopy

Hot conditioning operation of the primary heat transport system is an important step prior to the commissioning of Pressurized Heavy Water Reactors. One of the major objectives of the operation is to develop a stable and protective magnetite layer on the inner surfaces of carbon steel piping. The correlation between stable magnetite film growth on carbon steel surfaces and the period of exposure to hot conditioning environment is generally established by a combination of weight change measurements and microscopic/morphological observations of the specimens periodically removed during the operation. In the present study, electrochemical impedance spectroscopy (EIS) at room temperature is demonstrated as an alternate, quantitative technique to arrive at an optimal duration of the exposure period. Specimens of carbon steel were exposed for 24, 35 and 48 h during hot conditioning of primary heat transport system of two Indian PHWRs. The composition and morphology of oxide films grown during exposure was characterized by X-ray diffraction and optical microscopy. Further, ex situ electrochemical impedance spectra of magnetite films formed after each exposure were measured, in 1 ppm Li + electrolyte at room temperature as a function of potential in a range of −0.8 to +0.3 V SCE . The defect density of the magnetite films formed after each exposure was estimated by Mott–Schottky analysis of capacitances extracted from the impedance spectra. Further the ionic resistance of the oxide was also extracted from the impedance spectra. Defect density was observed to decrease with increase in exposure time and to saturate after 35 h, indicating stabilisation of the barrier layer part of the magnetite film. The values of the ionic transport resistance start to increase after 35–40 h of exposure. The quantitative ability of EIS technique to assess the film quality demonstrates that it can be used as a supplementary tool to the thickness and morphological characterizations of samples during hot conditioning.

Optimization of stress relief heat treatment of PHWR pressure tubes (Zr–2.5Nb alloy)

The micro-structure of cold worked Zr–2.5%Nb pressure tube material consists of elongated grains of α-zirconium enclosed by a thin film of β-zirconium phase. This β-Zr phase is unstable and on heating, progressively decomposes to α-Zr phase and β-phase enriched with Nb and ultimately form β Nb. Meta-stable ω-phase precipitates as an intermediate step during decomposition depending on the heat treatment schedule, β Zr → α + β enr → α + ω + β enr → α + β enr → α + β Nb Morphological changes occur in the β-zirconium phase during the decomposition. The continuous ligaments of β Zr phase turn into a discontinuous array of particles followed by globulization of the β-phase. The morphological changes impose a significant effect on the creep rate and on the delayed hydride cracking velocity due to reduction in the hydrogen diffusion coefficient in α Zr. If the continuity of β-phase is disrupted by heat treatment, the effective diffusion coefficient decreases with a concomitant reduction in DHC velocity. The pressure tubes for the Indian PHWRs are made by a process of hot extrusion followed by cold pilgering in two stages and an intermediate annealing. Autoclaving at 400 °C for 36 h ensures stress relieving of the finished tubes. In the present studies, autoclaving duration at 400 °C was varied from 24 h to 96 h at 12 h-steps and the micro-structural changes in the β-phase were observed by TEM. Dislocation density, hardness and the micro-structural features such as thickness of β-phase, inter-particle spacing and volume fraction of the phases were measured at each stage. Autoclaving for a longer duration was found to change the morphology of β-phase and increase the inter-particle spacing. Progressive changes in the aspect ratio of the β-phase and their size and distribution are documented and reported. These micro-structural modifications are expected to decrease DHC velocity during reactor operation.

Fitness for service assessment of coolant channels of Indian PHWRs

A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr–2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

Post irradiation examination of thermal reactor fuels

The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (∼15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ∼11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

Diagnostic system for identification of accident scenarios in nuclear power plants using artificial neural networks

This paper presents the work carried out towards developing a diagnostic system for the identification of accident scenarios in 220 MWe Indian PHWRs. The objective of this study is to develop a methodology based on artificial neural networks (ANNs), which assists in identifying a transient quickly and suggests the operator to initiate the corrective actions during abnormal operations of the reactor. An operator support system, known as symptom-based diagnostic system (SBDS), has been developed using ANN that diagnoses the transients based on reactor process parameters, and continuously displays the status of the reactor. As a pilot study, the large break loss of coolant accident (LOCA) with and without the emergency core cooling system (ECCS) in reactor headers has been considered. Several break scenarios of large break LOCA have been analyzed. The time-dependent transient data have been generated using the RELAP5 thermal hydraulic code assuming an equilibrium core, which conforms to a realistic estimation. The diagnostic results obtained from the ANN study are satisfactory. These results have been incorporated in the SBDS software for operator assistance. A few important outputs of the SBDS have been discussed in this paper.

Simulation of pressure tube deformation during high temperature transients

In the case of a loss of coolant accident (LOCA) with/without loss of emergency core cooling system (ECCS), in a PHWR, the pressure tube temperature could rise significantly depending on the scenario due to fuel heat up, with the system internal pressure also varying from 9 to 0.1 MPa during the event. This transient can cause metallurgical and geometrical changes in the pressure tube. Depending on the temperature reached, its structural integrity needs to be ensured during such events with the possibility of pressure tube–Calandria Tube contact. At high temperature the pressure tube deforms plastically due to internal pressure and fuel weight. Both ballooning and sagging behaviour are observed in the pressure tube. However, the relative contribution of each deformation is dictated by the pressure and temperature transient. Pressure tube sagging would always lead to an initial bottom contact. In such cases, studies have shown that the temperature difference between the top and bottom of the pressure tube could be as large as 500 °C. Along with the internal pressure, the thermal stress developed may threaten the integrity of the pressure tube. Modelling has been carried out to simulate the simultaneous sagging and ballooning deformation behaviour expected during the channel heat up condition. This paper describes the prediction of creep deformation in the context of the pressure tube configuration in Indian PHWR's and the study of the simultaneous ballooning and sagging behaviour of a pressure tube. The paper also presents the application of this model for a specific postulated LOCA in Indian PHWRs.

A new method for detecting pressure tube failures in Indian PHWRs

For the annulus gas system (AGS) of the standardised Indian pressurised heavy water reactor, an elaborate pressure tube (PT) crack monitoring and detection system is envisaged to ensure safety through leak-before-break. The parameters that are monitored relate to the detection of D 2O moisture leaking in from the primary heat transport (PHT) system through a cracked PT. Since a slow build-up of moisture in the AGS may also occur for reasons other than PT failure, it is desirable that a diverse measurement technique should be available. This paper suggests such a technique, based on the observation that a small reference concentration of fission gases is normally present in the annulus gas. This concentration would change sharply upon PT failure, when the heavy water from the leaking PHT system releases the dissolved fission gas content into the annulus. This paper presents a theoretical study of the parameters that influence the build-up of fission product noble gases in the AGS and shows that leakage rates as low as 10 g h −1 from a PT crack can be detected in a few tens of minutes by this method. This is expected to substantially increase the available time between the leak detection and the PT failure, thus serving as an important tool in meeting the leak-before-break criterion of a critical component in PHWRs.

Structural evolution of containment for indian Phwrs

Pressurised heavy water reactors form the mainstay of the first stage of the Indian Nuclear Power Programme. Each of these units is provided with containment as an ultimate barrier for escape of radioactivity between the inside of the reactor building and the atmosphere. These containments have thus the function of limiting leakage of gases from inside the reactor building to the external atmosphere under the most adverse accidental conditions. To achieve this, beside proper design and construction of containments, suitable energy management features are also incorporated to minimise the pressure and temperature buildup during an accidental condition. Present paper deals with the step by step evolution of containments for Indian PHWRs. This starts from the use of single containment with energy management system based on active components such as dousing system, to a complete double containment with passive energy management system in the form of vapour suppression pool. This evolution is based on design criteria and functional requirements as well as various research and developmental work done in different areas related to containment design. The text initiates the discussion with the brief description of containment systems for different power reactors in India. This is followed by the design criteria and the functional requirements adopted for the containments. Identification of different areas chosen for developmental work and their influence on the design/construction of our containments are described next. It concludes with the indication of future thrust in this area.