Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

Science.gov (United States)

Saibaba, N.

2008-12-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties.

Fabrication of seamless calandria tubes by cold pilgering route using 3-pass and 2-pass schedules

International Nuclear Information System (INIS)

Saibaba, N.

2008-01-01

Calandria tube is a large diameter, extremely thin walled zirconium alloy tube which has diameter to wall thickness ratio as high as 90-95. Such tubes are conventionally produced by the 'welded route', which involves extrusion of slabs followed by a series of hot and cold rolling passes, intermediate anneals, press forming of sheets into circular shape and closing the gap by TIG welding. Though pilgering is a well established process for the fabrication of seamless tubes, production of extremely thin walled tubes offers several challenges during pilgering. Nuclear fuel complex (NFC), Hyderabad, has successfully developed a process for the production of Zircaloy-4 calandria tubes by adopting the 'seamless route' which involves hot extrusion of mother blanks followed by three-pass pilgering or two-pass pilgering schedules. This paper deals with standardization of the seamless route processes for fabrication of calandria tubes, comparison between the tubes produced by 2-pass and 3-pass pilgering schedules, role of ultrasonic test charts for control of process parameters, development of new testing methods for burst testing and other properties

Design of calandria-end shield support diaphragm of Narora Atomic Power Project

Energy Technology Data Exchange (ETDEWEB)

Srivastava, S K; Nanda Kumar, S; Kakodkar, A

1975-01-01

The calandria-end shield diaphragm is one of the important components in Narora Atomic Power Plant. The support diaphragm is designed against elastic and plastic instability failures. Method of analysis for elastic and plastic instability is discussed for normal loading, pipe rupture loading, and earthquake loading.

Design of calandria-end shield support diaphragm of Narora Atomic Power Project

International Nuclear Information System (INIS)

Srivastava, S.K.; Nanda Kumar, S.; Kakodkar, A.

1975-01-01

The calandria-end shield diaphragm is one of the important components in Narora Atomic Power Plant. The support diaphragm is designed against elastic and plastic instability failures. Method of analysis for elastic and plastic instability is discussed for normal loading, pipe rupture loading and earthquake loading. (author)

Performance assessment of turbulence models for the prediction of moderator thermal flow inside CANDU calandria

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong

2012-01-01

The moderator thermal flow in the CANDU calandria is generally complex and highly turbulent because of the interaction of the buoyancy force with the inlet jet inertia. In this study, the prediction performance of turbulence models for the accurate analysis of the moderator thermal flow are assessed by comparing the results calculated with various types of turbulence models in the commercial flow solver FLUENT with experimental data for the test vessel at Sheridan Park Engineering Laboratory (SPEL). Through this comparative study of turbulence models, it is concluded that turbulence models that include the source term to consider the effects of buoyancy on the turbulent flow should be used for the reliable prediction of the moderator thermal flow inside the CANDU calandria

Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

International Nuclear Information System (INIS)

Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

1983-01-01

A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

International Nuclear Information System (INIS)

Abucay, R.C.; Mahil, K.S.; Goszczynski, J.J.

1995-01-01

The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

Development of video probe system for inspection of feeder pipe support in calandria reactor

International Nuclear Information System (INIS)

Cho, Jai Wan; Lee, Nam Ho; Choi, Young Soo

2000-07-01

There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post- Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And untrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughlv because of narrow and confined accessibility, that is, an inspection space between the pressure tube channels is less than 100mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feederpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant

Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface

Energy Technology Data Exchange (ETDEWEB)

Behdadi, Azin, E-mail: behdada@mcmaster.ca; Talebi, Farshad; Luxat, John

2017-04-15

Highlights: • Pressure tube-calandria tube contact may challenge fuel channel integrity in CANDU. • Critical heat flux variation is predicted on the outer surface of CANDU calandria tube. • A two-phase boundary layer flow driven by buoyancy is modeled on the surface. • Different slip ratios and flow regimes are considered inside the boundary layer. • Subcooling effects are added to the model using wall heat flux partitioning. - Abstract: One accident scenario in CANDU reactors that can challenge the integrity of the primary pressure boundary is a loss of coolant accident, referred to as critical break LOCA, in which the pressure tube (PT) can undergo thermal creep strain deformation and contact its calandria tube (CT). In such case, rapid redistribution of stored heat from PT to CT, leads to a large spike in heat flux to the moderator which can cause bubble accumulation and dryout on the CT surface. A challenge to fuel channel integrity is posed if critical heat flux occurs on the surface of the CT and results in sustained film boiling. If the post-dryout temperature becomes sufficiently high then continued creep strain of the PT and CT may lead to fuel channel failure. In this study, a mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of CT. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow. The model shows good agreement with the available experimental data and has been modified to take into account the effect of subcooling.

Improving accuracy of ET measurement of LISS nozzle to calandria tube clearance

International Nuclear Information System (INIS)

Craig, S.T.; Krause, T.W.; Schankula, J.J.

2006-01-01

The AECL Fuel Channel Inspection System (AFCIS) has been used in an in-reactor field trial to successfully measure the clearance between Liquid Injection Shutdown System (LISS) nozzles and calandria tubes. Each measurement over the full length of a channel added only 15 minutes to the on-channel inspection time. No changes were required to the inspection heads. The only equipment changes made were the addition of a Remote Field Eddy Current (RFEC) module to the eddy current instrument, and minor wiring changes, at the instrument, to achieve a RFEC configuration. With the experience gained from the field trial, factors potentially limiting accuracy were identified. These, and other factors, were investigated and are discussed herein. The RFEC probe is delivered inside the pressure tube. Magnetic fields from the RFEC probe extend through the conducting walls of the pressure tube and calandria tube to interact with the LISS nozzle. Data acquired during the field trial showed the LISS nozzle signal is distinct and the signal-to-noise ratio is very favourable. Nevertheless, comparison of the RFEC measurements to a visual examination, made during the same outage, had the RFEC method underestimating the clearance by 2.5 mm on average. By way of laboratory tests, the following factors were investigated as potential sources of error: resistivity and geometry of LISS nozzle reference/calibration pieces, pressure-tube wall thickness, diameter and resistivity variations, pressure-tube to calandria-tube gap, and radial offsets of the probe within the pressure-tube. The sensitivity to these various noise sources was established. A model, based on fundamental electromagnetic principles, was developed and was used to normalize the effects of LISS nozzle conductivity and geometry. This enabled compensation for various sources of error, and made it possible to produce a correction factor for the field trial data, reducing the average difference from the visual inspection of LISS

The thermal interaction of a buoyant plume from a calandria tube with an oblique jet

Energy Technology Data Exchange (ETDEWEB)

Rossouw, D.J.; Atkins, M.D.; Beharie, K. [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: tong.kim@wits.ac.za [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejun (Korea, Republic of)

2016-12-15

Highlights: • A crucial role of relative orientation between mixed convection modes is observed. • The extent of thermal interaction strongly depends on the relative orientation. • Coolant flow is substantially diffused by a buoyant plume if counter-acting. • Slightly oblique coolant flow to the gravitational axis provides the best cooling. - Abstract: Severe reactor core damage may occur from fuel channel failure as a consequence of excessive heat emitted from calandria tubes (CTs) in a pressurised heavy water (D{sub 2}O) reactor (CANDU). The heating of the CTs is caused by creep deformation of the pressure tubes (PTs), which may be ballooning or sagging depending on the internal pressure of the PTs. The deformation of the pressure tube is due to overheating as a result of a loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) failure. To prevent the exacerbation of the LOCA, circulating D{sub 2}O in the moderator tank may be utilized by forming a secondary jet that externally cools the individual CTs. The buoyant plume develops around the CTs and interacts with the secondary jet at a certain oblique angle with respect to the gravitational axis, depending on the spatial location of the hot calandria tubes (or the hot reactor core region). This study reports on how the local and overall heat transfer characteristics on a calandria tube where the buoyant plume develops, are altered by the obliqueness of the external secondary jet (from a co-current jet to a counter-current jet) in a simplified configuration at the jet Reynolds number of Re{sub j} = 1500 for the Archimedes number of Ar{sub D} = 0.11 and Rayleigh number of Ra{sub D} = 1.6 × 10{sup 6} (modified Rayleigh number of 3.0 × 10{sup 7}).

Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

2015-04-01

Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

Visualization of Moderator Circulation with 1/8 scaled HU-KINS for Calandria Tank of CANDU6

International Nuclear Information System (INIS)

Cho, Dae Kun; Kim, Nam Seok; Lee, Jae Young; Kim, Man Woong

2005-01-01

The moderator cooling capability of Calandria tank is important during the hypothetical incident of CANDU reactor. Even though the loss of ECCS cooling, the heat of the nuclear fuel can be cooled by the heat transfer path through the contacting location between the pressure tube and Calandria tube. The moderator cooling system removes the heat from the fuel so it pays an ultimate heat sink for the nuclear safety. Therefore, the understanding and estimating the thermal hydraulic conditions of the circulating moderator in the Calandria is of importance. However, the three dimensional nature and complexity of the Calandria tubes make it difficult to analyze. Recent progress in the the three dimensional computational fluid dynamics including the model of turbulent flow stimulate the studies on this subject. For the design or safety analysis purpose, the mesh generation and furnishing proper physical models are imperative and the experimental validation should be performed. However, the experimental facilities have been constructed in the scaled way by the Canadian research groups, AECL and COG. Two facilities: SPEL (1/10 scale by Koroyannaski, 1983) and STERN (1/4 scale by Hadaller, 1990) have been producing experimental data such as the local temperatures and velocities. Several papers has been published to report their CFD codes can be available for the CANDU analysis by comparing the calculations with the experimental results of those facilities. However, as noted by Lee et al. the previous experimental facilities were not scaled properly both in the sense of the force balance between buoyancy force and the jet inertia force. Also, the power density of the moderator was not scaled properly. Therefore, it cannot be said that the observations through SPEL and STERN may have a certain discrepancy from the real CANDU-6 plant. Lee et al (2003) developed the scaling laws for the CANDU-6 and design a 1/8 scale experimental facility named as HU-KINS. In the present paper

Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

International Nuclear Information System (INIS)

Luxat, J.C.

2002-01-01

Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

The transition criteria of circulating flow pattern of moderator in the calandria tank of CANDU nuclear power plant

International Nuclear Information System (INIS)

Jung, Yun Sik; Lee, Jae Young; Kim, Man Woong

2004-01-01

The moderator cooling system to the Calandria tank of CANDU nuclear power plant provides an alternative pass of heat sink during the hypothetical loss of coolant accident. Also, the neutron population in the CANDU plant can be affected by the moderator temperature change which strongly depends on the circulating flow pattern in the Calandria tank. It has been known that there are three distinguished flow patterns: the buoyancy dominated flow, the momentum dominated flow, and the mixed type flow. The Canadian Nuclear Safety Commission (CNSC) recommended that a series of experimental works should be performed to verify the three dimensional codes. Two existing facilities, SPEL (1982) and STERN (1990), have produced experimental data for these purposes. The present work is also motivated to build up a new scaled experimental facility named HGU for the same purposes. CANDU-6 was selected as the target plant to be scaled down. In the design for the scaled facility, the knowledge on the flow regime transitions in the circulating flow was imperative. In the present study, to pave the way for the scaling, the flow pattern maps of circulating flow were constructed based on the Reynolds number and Archimedes number. The CFX code was employed with real meshes to represent all calandria tubes in the tank. The flow pattern maps were constructed for SPEL, STERN, HGU, and CANDU6. As the key transition criterion useful for scaling law, a new Archimedes number considering the jet impingement of the feed water in the Calandria tank was found. The transition of flow patterns was made with the same Archimedes number for CANDU6, STERN and HGU. However, SPEL which has third of the modified Archimedes number showed different maps in the wider region of mixed flow pattern was observed. It was found that the Archimedes number considering the inlet nozzle velocity plays the key role in patterns classification. Also, it can be suggested that the moderator cooling system needs to be designed

Remote tooling for inspection and repair in Pickering NGS-A calandria vault

International Nuclear Information System (INIS)

Hadji-Mirzai, M.; Tokarz, A.; Vandenberg, J.P.

1993-01-01

In recent years it has been necessary to develop capabilities for the inspection and repair of carbon steel components located within calandria vaults at Ontario Hydro's Pickering Nuclear Generating Station 'A'. Concerns about corrosion of piping and some of the structural components have made necessary the development of remote manipulators to inspect and repair carbon steel components within the vaults to ensure continued reliable operation of the units. Remote manipulators for this program have been designed to perform a number of inspection and repair tasks, and several versions have been developed to specialise in detailed inspection techniques and precision tooling module manipulation. (author)

2D modeling of moderator flow and temperature distribution around a single channel after pressure tube/calandria tube contact

International Nuclear Information System (INIS)

Behdadi, A.; Luxat, J.C.

2009-01-01

A 2D computational fluid dynamics (CFD) model has been developed to calculate the moderator velocity field and temperature distribution around a single channel inside the moderator of a CANDU reactor after a postulated ballooning deformation of the pressure tube (PT) into contact with the calandria tube (CT). Following contact between the hot PT and the relatively cold CT, there is a spike in heat flux to the moderator surrounding the CT which may lead to sustained CT dryout. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The present research is focused on establishing the limits for dryout occurrence on the CTs for the situation in which pressure tube-calandria tube contact occurs. In order to consider different location of the channels inside the calandria, both upward and downward flow directions have been analyzed. The standard κ - ε turbulence model associated with logarithmic wall function is applied to predict the effects of turbulence. The governing equations are solved by the finite element software package COMSOL. The buoyancy driven natural convection on the outer surface of a CT has been analyzed to predict the flow and temperature distribution around the single CT considering the local moderator subcooling, wall temperature and heat flux. The model also shows the effect of high CT temperature on the flow and subcooling around the CTs at higher/lower elevation depending on the flow direction in the domain. According to the flow pattern and temperature distribution, it is predicted that stable film boiling generates in the stagnation region on the cylinder. (author)

New contact boiling experiments to evaluate Calandria tube strain acceptance criteria

Energy Technology Data Exchange (ETDEWEB)

El-Hawary, M.; Szymanski, J.; Tanase, A.; Delja, A.; Oussoren, A., E-mail: Magdy.El-Hawary@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Comission, Ottawa, ON (Canada); Neal, P. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

2015-07-01

The Canadian Nuclear Safety Commission(CNSC) has contracted the Canadian Nuclear Laboratories(CNL) to conduct additional Contact Boiling (CB) experiments with the main objective of evaluating the acceptance criterion of CalandriaTube (CT) strain limit of 2%, proposed by the industry for fuel channel integrity assessments. The test conditions are selected using analytical tools and guidance from existing CANDU Owners Group (COG) test results, so as to lead to CT strain close to this value. The experiments will also be used to evaluate the CT quench temperature correlation proposed. This paper presents conditions selected for the first three experiments, their most important results and their preliminary analysis, with a focus on the test which produced CT strain in excess of 2%. (author)

Visual inspection technology of the narrow and small confined area for monitoring feederpipe support of pressure tube in calandria reactor

Energy Technology Data Exchange (ETDEWEB)

Cho, Jae Wan; Lee, Nam Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feedeerpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant. 45 figs.,31 tabs. (Author)

Mixed convection around calandria tubes in a ¼ scale CANDU-6 moderator circulation tank

Energy Technology Data Exchange (ETDEWEB)

Atkins, M.D.; Rossouw, D.J.; Boer, M. [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: tong.kim@wits.ac.za [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-05-15

Highlights: • A secondary jet is formed at a stagnation region and is directed towards the center of the MCT. • The secondary jet undergoes the significant dissipation and mixing due to calandria tubes (CTs). • Its cooling effectiveness is reduced on the CTs in the bottom of the MCT. • With forced convection dominance, peak heat transfer is on the upper CT surface. • With natural convection dominance, peak heat transfer is on the lower CT surface. - Abstract: This study experimentally characterizes mixed convection around calandria tubes (CTs) in a ¼ scale CANDU-6 moderator circulation tank (MCT) that uses air as the working fluid. In a full scale CANDU-6 reactor that undergoes a postulated dual failure with a loss-of-coolant accident without the emergency core cooling system available, mixed convection heat transfer occurs around the CTs. The cooling effectiveness of the moderator is diminished as an emergency heat sink if overheating eventually leads to film boiling. To prevent the onset of film boiling, local sub-cooling margins of the moderator needs to be maintained or else the critical heat flux should be increased. Circulating the moderator which interacts with the overheated CTs increases the heat transfer into the moderator which may suppress film boiling. The present experimental results demonstrate that the cooling effectiveness of the circulating moderator, in particular the secondary jet, is attenuated substantially as it is convected away from the inner wall towards the center of the MCT. The momentum of the secondary jet is diffused through the CTs. At a low jet Reynolds number, the secondary jet becomes ineffective so that some overheated CTs positioned in the other half of the MCT are cooled only by natural convection.

Experimental investigation of coolant and poisoned moderator mixing due to a simulated pressure tube/calandria tube fishmouth rupturing an overpoisoned guaranteed shutdown state

International Nuclear Information System (INIS)

Mackinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

1997-01-01

During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The contents of the paper are as follows. First, the objectives of the experimental program are

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

International Nuclear Information System (INIS)

Jiang, J.T.; Luxat, J.C.

2008-01-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

2008-07-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

In reactor measurements, modeling and assessments to predict liquid injection shutdown system nozzle to Calandria tube time to contact

International Nuclear Information System (INIS)

Kirstein, K.; Kalenchuk, D.

2011-01-01

Over the past few years there has been an expanding effort to assess the potential for Calandria Tubes (CTs) coming into contact with Liquid Injection Shutdown System (LISS) Nozzles to ensure continued contact-free operation as required by CSA N285.4. LISS Nozzles (LINs), which run perpendicular to and between rows of fuel channels, sag at a slower rate than the fuel channels. As a result certain LINs may come in contact with CTs above them. The CT/LIN gaps can be predicted from calculated CT sag, LIN sag and a number of component and installation tolerances. This method however results in very conservative predictions when compared to measurements, confirmed with the in reactor measurements initiated in 2000, when gaps were successfully measured the first time using images obtained from a camera-assisted measurement tool inserted into the calandria. To reduce the conservatism of the CT/LIN gap predictions, statistical CT/LIN gap models are used instead. They are derived from a comparison between calculated gaps based on nominal dimensions and the visual image based measured gaps. These reactor specific (typically 95% confidence level) CT/LIN gap models account for all uncertainties and deviations from nominal values. Prediction error margins reduce as more in-reactor gap measurements become available. Each year more measurements are being made using this standardized visual CT/LIN proximity method. The subsequently prepared reactor-specific models have been used to provide time to contact for every channel above the LINs at these stations. In a number of cases it has been used to demonstrate that the reactor can be operated to its end of life before refurbishment with no predicted contact, or specific at-risk channels have been identified for which appropriate remedial actions could be implemented in a planned manner. (author)

Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

2005-01-01

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T.; Luxat, J.C. [McMaster University, A315 JHE Building, 1280 Main St.W. Hamilton, ON, L8S 4L7 (Canada)

2008-07-01

Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

International Nuclear Information System (INIS)

Jiang, J.T.; Luxat, J.C.

2008-01-01

Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

Methods of evaluation of accuracy with multiple essential parameters for eddy current measurement of pressure tube to calandria tube gap in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Shokralla, S., E-mail: shaddy.shokralla@opg.com [Ontario Power Generation, IMS NDE Projects, Ajax, Ontario (Canada); Krause, T.W., E-mail: thomas.krause@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

2014-01-15

The purpose of inspection qualification of a particular inspection system is to show that it meets applicable inspection specification requirements. Often a requirement of the inspection system is that it meets a particular accuracy. In the case of a system with multiple inputs accompanied by additional influential parameters, calculation of the system's output accuracy can be formidable. Measurement of pressure-tube to calandria tube gap in CANDU reactors using an eddy current based technique is presented as a particular example of a system where multiple essential parameters combine to generate a final uncertainty for the inspection system. This paper outlines two possible methods of calculating such a system's accuracy, and discusses the advantages and disadvantages of each. (author)

The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

Directory of Open Access Journals (Sweden)

Hyoung Tae Kim

2016-01-01

Full Text Available The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor, has been modeled in multidimension for the computation based on CFD (computational fluid dynamics technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

Computer modelling of eddy current probes for ISI of pressure tube/calandria tube assemblies in PHWRs

International Nuclear Information System (INIS)

Rao, B.P.C.; Shyamsunder, M.T.; Bhattacharya, D.K.; Raj, Baldev

1992-01-01

Non-destructive Evaluation (NDE) plays a major role in ensuring the safe and reliable operation of PHWRs which are the mainstay of India's nuclear power programme. An important in-service inspection (ISI) requirement in these reactors is carried out through Eddy Current Testing (ECT) of the pressure tube (PT)/calandria tube (CT) assemblies. The material of construction of these assemblies is zircaloy-2. The two main objectives of this ISI are the detection of garter spring between CT and PT and the profiling of gap between CT and PT. The paper discusses the work carried out at the authors' laboratory on the development of ECT probes for ISI of PT/CT assemblies. Emphasis has been given on the work done on the design and optimisation of the probes using computer modeling. A 2-D finite element code has been developed for this purpose. The code is developed around a diffusion equation which can be derived from Maxwell's equations governing the electromagnetic phenomenon. An axisymmetry has been considered, since the probes are bobbin type. Results of impedance plane outputs obtained by modelling and those by experiments using actual probes have shown good matching. Salient features of an indigenously developed interactive PC based data acquisition, analysis and retrieval system to cater to ISI of PC/CT assemblies are described. (author). 10 refs., 7 figs

The influence of the preliminary garter spring spacer simulator clamping force in the pressure tube spacer -calandria tube hook-up simulator aging behaviour

International Nuclear Information System (INIS)

Gyongyosi, T.; Deloreanu, G.; Puiu, D.; Corbescu, B.; Anghel, N.; Dinu, E.

2016-01-01

The garter spring spacer is a specially constructed torsion spring used to fit-out the CANDU 6 fuel channel. The pressure tube ageing decreases the gap to the calandria tube. Continuous gap decrease directly affects the garter spring spacers behavior during fuel channel assembly operation. The preliminary clamping force value of the garter spring spacer assembly is important for its ageing behavior. This paper briefly describes the experimental technological facilities used for conducted the experiments and highlights some of the important moments during an experiment carried out in laboratory conditions, without using pressurized boiled water and irradiation working conditions. The results analysis and some conclusions are outlined at the end, pointing out that a garter spring spacer preliminary clamping force increase reduces the vibration response signal amplitude, and does not lead to its relaxation. The paper is dedicated to specialists working in research and technological engineering. (authors)

Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

International Nuclear Information System (INIS)

Kim, Tae Ryong; Sohn, Seok Man; Lee, Jun Shin; Lee, Sun Ki; Lee, Jong Po

2001-01-01

Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

External glass peening of zircaloy calandria tubes to increase the critical heat flux

International Nuclear Information System (INIS)

Fong, R.W.L.; Coleman, C.E.; Nitheanandan, T.; Kroeger, V.D.; Moyer, R.G.; Sanderson, D.B.; Root, J.H.; Rogge, R.B.

1997-12-01

Glass-peening the outside surfaces of Zircaloy calandria tubes increases the nucleation sites available for boiling heat transfer and has been demonstrated to enhance the critical heat flux (CHF) in pool-boiling experiments. The objective of this study is to optimise the heat-transfer enhancement by glass peening while ensuring that the microstructure of the peened tube is acceptable for reactor use. Pool-boiling tests were done using small Zircaloy tubes with as-received ('smooth') surfaces and variously peened surfaces, to evaluate two peening parameters, glass-bead size and the coverage of peened surface. Our results showed that the maximum enhancement of CHF (by 60% compared with as-received tubes) was obtained using a glass-bead size of 90-125 μm with a coverage of 100%. The CHF enhancement was found to be insensitive to glass-bead size over a wide range (from 60-90 μm to 125-180 μm). Using a fixed glass-bead size of 125-180 μm to evaluate the influence of peening coverage, the maximum effect on the CHF response was obtained with a coverage of 1 00%. The microstructures of the peened tubes were evaluated using light microscopy, X-ray and neutron diffraction, and mechanical tests. After peening, the microstructure in the subsurface layer (-30 μm) consisted of deformed α-Zr grains, and the crystallographic texture of the grains changed slightly. After stress-relieving at 500 degrees C for 1 h, some recrystallisation had occurred and the residual strains remaining in the tube were low. The tensile and burst properties of glass-peened and stress-relieved tubes were similar to those of as-received tubes. The microstructures introduced by peening and stress relieving were judged to have little effect on creep and growth behaviour. Since there are no deleterious consequences of the glass-peening treatment, the peened and stress-relieved tubes are found to be acceptable for reactor use. (author)

Prediction of moderator temperature under 35% RIH break LOCA with LOECC in CANDU calandria vessel

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung; Lee, Jae Yung

2004-01-01

A CANDU reactor has the unique safety features with the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors such as a PWR. One of the safety features is that the heavy water moderator is continuously cooled, providing with a heat sink for the decay heat produced in the fuel when there is the LOCA with the coincident failure of the emergency coolant injection (ECI) system. Under such a dual failure condition, the hot pressure tube (PT) would deform into contacting with the calandria tube (CT), providing with an effective heat transfer path from the fuel to the moderator. Following PT/CT contact, there is the spike of the heat flux in the moderator surrounding the CT, which could lead to sustained CT dryout. The prevention of the CT dryout depends on available local moderator subcooling. Higher moderator temperature (or lower subcooling) would decrease the margin of the CTs to dryout. As for LOCAs with coincident loss of the ECI, fuel channel integrity depends on the capability of the moderator as an ultimate heat sink. In this regard, the Canadian Nuclear Safety Commission (CNSC) had categorized the temperature prediction for the moderator cooling integrity as a general action item (GAI) and had recommended that a series of experimental works should be performed to verify the evaluation codes comparing with the results of three-dimensional experimental data. However, although a couple of computer codes were used to predict moderator temperature prediction for those problems, they could not be adequately validated due to the uncertainty of temperature prediction. In this work, the temperature prediction under the transient condition of LOCA with loss of emergency core cooling (LOECC) in a CANDU reactor is conducted using the optimized calculation scheme from the previous work

Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

International Nuclear Information System (INIS)

Park, S. Y.; Song, Y. M.

2015-01-01

The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building

Severe accident development modeling and evaluation for CANDU

Energy Technology Data Exchange (ETDEWEB)

Negut, Gheorghe [National Agency for Radioactive Waste, 1, Campului Str., 115400 Mioveni (Romania)], E-mail: gheorghe.negut@andrad.ro; Catana, Alexandru [Institute for Nuclear Research Pitesti, 1, Campului Str., Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, Ilie; Dupleac, Daniel [Politehnica University Bucharest, 313, Splaiul Independentei, Sect. 6, 060042 Bucharest (Romania)

2009-09-15

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

Severe accident development modeling and evaluation for CANDU

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2009-01-01

Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents. Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data.

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

CANDU severe accident analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

IAEA ICSP on HWR moderator subcooling requirements to demonstrate backup heat sink

International Nuclear Information System (INIS)

Choi, J.; Nitheanandan, T.

2013-01-01

The IAEA launched a new International Collaborative Standard Problem (ICSP) on 'HWR Moderator Subcooling Requirements to Demonstrate Backup Heat Sink Capabilities of Moderator during Accidents'. The purpose of the ICSP is to benchmark analysis computer codes in simulating contact boiling experimental data to assess the subcooling requirements for an overheated pressure tube, plastically deforming into contact with the calandria tube during a postulated large break loss of coolant accident. The experimental data obtained for the ICSP blind simulation can be used to assess safety analysis computer codes simulating thermal radiation heat transfer to the pressure tube, pressure tube deformation or failure, pressure tube to calandria tube heat transfer, calandria tube to moderator heat transfer, and calandria tube deformation or failure. (author)

Pressure tube reactor

International Nuclear Information System (INIS)

Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

1993-01-01

In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

Flooding of a large, passive, pressure-tube LWR

Energy Technology Data Exchange (ETDEWEB)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1995-09-01

A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

Energy Technology Data Exchange (ETDEWEB)

De, T K; Collins, W M; Holmes, R W [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

CANDU nuclear reactors use D{sub 2}0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs.

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

International Nuclear Information System (INIS)

De, T.K.; Collins, W.M.; Holmes, R.W.

1995-01-01

CANDU nuclear reactors use D 2 0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs

A CANDU Severe Accident Analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

2006-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

Thermal hydraulics of CANDU severe accident analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

Design and development of rolled joint for moderator sparger channel of an Indian Pressurised Heavy Water Reactor

International Nuclear Information System (INIS)

Joemon, V.; Sinha, R.K.

1993-01-01

Indian Pressurised Heavy Water Reactors are natural uranium fuelled heavy water moderated and cooled reactors. As per the conventional scheme, the moderator enters through one or more inlet nozzles penetrating the calandria shell and flows out through outlet nozzles. Baffles are fixed at the inlet nozzles for proper distribution of moderator in the calandria and to avoid the impact of the jet on the neighbouring calandria tubes. An alternate scheme for moderator inlet has been conceived and engineered in which three lower peripheral lattice locations of the reactor are converted into moderator inlets. This is achieved by moderator sparger channels each containing a 5 m long perforated zircaloy-2 sparger tube rolled to the calandria tube sheets and extended by stainless steel tubular components (inserts) at both ends of a sparger channel. Moderator enters the sparger channel at both ends and flows into the calandria. In the absence of standard codes for design of rolled joints, it was requires to develop these joints based on trials followed by various tests. this paper discusses the details of the rolled joint developed for this purpose, the details of the trials with test results and optimization of rolling parameters for these joints

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

Khater, H.A.; Hadaller, G.I.; Stern, F.

1985-06-01

A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

Energy Technology Data Exchange (ETDEWEB)

Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

2015-10-15

Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as fuel channel contact experiments. The difference between available subcooling and required subcooling is called subcooling margins. The moderator flow circulation patterns are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep

KAPP-3 and 4 containment pressure following postulated severe accident along with SAMG implementation

International Nuclear Information System (INIS)

Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, Nalini; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

2011-01-01

Containment is an ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following severe accident i.e. Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet header with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are assumed to be added to calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. After moderator and calandria vault water have completely boiled off, rapid hydrogen generation would take place due to oxidation of pressure tubes and calandria tubes. In such accident scenario, the core is severely damaged. It will also lead to release of a large quantity of radio nuclides to containment atmosphere. To arrest the progression of accident, which can result in Severe Core damage and large amount of hydrogen production, which could leads to containment failure due to hydrogen deflagration or detonation, application of Severe Accident Management Guidelines (SAMG) has been studied. SAMG involve addition of water to calandria and calandria vault. It would result the boiling of the added water and consequent pressurization of containment. This paper presents the analysis for pressure-temperature of KAPP-3 and 4 containment following the postulated accident along with the application of Severe Accident Management Guidelines (SAMG). SAMG initiated action helps in arresting the progression of core

Studies on flow induced vibration of reactivity devices of 700 MWe Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Prabhakaran, K.M., E-mail: kmprabha@yahoo.com [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Goyal, P.; Dutta, Anu; Bhasin, V.; Vaze, K.K.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Pillai, Ajith V.; Mathew, Jimmy [Nuclear Power Corporation of India Ltd., Mumbai 400 094 (India)

2012-03-15

Highlights: Black-Right-Pointing-Pointer FIV studies on internals of heavy water filled calandria of 700 MWe Indian PHWR is presented. Black-Right-Pointing-Pointer This includes CFD and structural dynamic analysis to predict the dynamic behavior of component lying inside calandria. Black-Right-Pointing-Pointer Results of these calculations as well as conclusions from this investigation are presented. Black-Right-Pointing-Pointer It is established that FIV is not a concern in the present design of calandria internals. - Abstract: Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

OpenFOAM Analysis of CANDU-6 Moderator Flow

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Chang, Se-Myong

2015-01-01

In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D 2 O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO 2 gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC C LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat transfer

OpenFOAM Analysis of CANDU-6 Moderator Flow

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, Se-Myong [Kunsan National University, Gunsan (Korea, Republic of)

2015-10-15

In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D{sub 2}O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO{sub 2} gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC{sub C}LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat

Process systems of PHWR - Indian experience

Energy Technology Data Exchange (ETDEWEB)

Ramandan, T S.V. [Madras Atomic Power Station (MAPS), Madras (India)

1991-04-01

Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

Process systems of PHWR - Indian experience

International Nuclear Information System (INIS)

Ramandan, T.S.V.

1991-01-01

Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

Falling film flow, heat transfer and breakdown on horizontal tubes

International Nuclear Information System (INIS)

Rogers, J.T.

1980-11-01

Knowledge of falling film flow and heat transfer characteristics on horizontal tubes is required in the assessment of certain CANDU reactor accident sequences for those CANDU reactors which use moderator dump as one of the shut-down mechanisms. In these reactors, subsequent cooling of the calandria tubes is provided by falling films produced by sprays. This report describes studies of falling film flow and heat transfer characteristics on horizontal tubes. Analyses using integral methods are given for laminar and turbulent flow, ignoring and accounting for momentum effects in the film. Preliminary experiments on film flow stability on horizontal tubes are described and various mechanisms of film breakdown are examined. The work described in this report shows that in LOCA with indefinitely delayed ECI in the NPD or Douglas Point (at 70 percent power) reactors, the falling films on the calandria tubes will not be disrupted by any of the mechanisms considered, provided that the pressure tubes do not sag onto the calandria tubes. However, should the pressure tubes sag onto the calandria tubes, film disruption will probably occur

Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

International Nuclear Information System (INIS)

Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

1998-08-01

The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

Operating experiences of reactor shutdown system at MAPS

International Nuclear Information System (INIS)

Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

1997-01-01

The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

Improvement of core degradation model in ISAAC

International Nuclear Information System (INIS)

Kim, Dong Ha; Kim, See Darl; Park, Soo Yong

2004-02-01

If water inventory in the fuel channels depletes and fuel rods are exposed to steam after uncover in the pressure tube, the decay heat generated from fuel rods is transferred to the pressure tube and to the calandria tube by radiation, and finally to the moderator in the calandria tank by conduction. During this process, the cladding will be heated first and ballooned when the fuel gap internal pressure exceeds the primary system pressure. The pressure tube will be also ballooned and will touch the calandria tube, increasing heat transfer rate to the moderator. Although these situation is not desirable, the fuel channel is expected to maintain its integrity as long as the calandria tube is submerged in the moderator, because the decay heat could be removed to the moderator through radiation and conduction. Therefore, loss of coolant and moderator inside and outside the channel may cause severe core damage including horizontal fuel channel sagging and finally loss of channel integrity. The sagged channels contact with the channels located below and lose their heat transfer area to the moderator. As the accident goes further, the disintegrated fuel channels will be heated up and relocated onto the bottom of the calandria tank. If the temperature of these relocated materials is high enough to attack the calandria tank, the calandria tank would fail and molten material would contact with the calandria vault water. Steam explosion and/or rapid steam generation from this interaction may threaten containment integrity. Though a detailed model is required to simulate the severe accident at CANDU plants, complexity of phenomena itself and inner structures as well as lack of experimental data forces to choose a simple but reasonable model as the first step. ISAAC 1.0 was developed to model the basic physicochemical phenomena during the severe accident progression. At present, ISAAC 2.0 is being developed for accident management guide development and strategy evaluation. In

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This volume of appendices presents listings and sample runs of the computer codes used in the study of the thermalhydraulic behaviour of CANDU reactor cores during severe loss of coolant accidents. The codes, written in standard FORTRAN, are MODBOIL, to calculate moderator temperatures, pressures and water levels; DEBRIS, to calculate the transient temperature distribution in the debris of calandria and pressure tubes and fuel pellets; MOLTENPOOL, to calculate the temperature history in a pool of molten debris; CONFILM, to calculate the behaviour of a condensing film of vaporized core debris on the calandria wall, and BLDG, to calculate the pressurization of the containment during the expulsion of moderator through pressure relief ducts. In addition there are discussions of the average condensation heat transfer coefficient for vaporized core material on the calandria wall, and of vapor explosions

Pressure tube reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Fujino, Michihira.

1980-01-01

Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

Safe new reactor for radionuclide production

International Nuclear Information System (INIS)

Gray, P.L.

1995-01-01

In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

2011-01-01

In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

Studies of loss-of-coolant and loss-of-regulation accidents

International Nuclear Information System (INIS)

Rogers, J.T.

1979-10-01

Studies of a CANDU reactor during loss of coolant with delayed emergency core cooling showed that the moderator is an effective heat sink, and that in reactors with moderator dump the calandria sprays provide effective cooling. Fuel channel melting would not occur, and a coolable geometry will be maintained. Studies on film cooling and film stability on calandria tubes and on the analysis of flow reversal in vertical feeder tubes are also reported

ACR-1000: Enhanced response to severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

International Nuclear Information System (INIS)

Brown, R.; Cobanoglu, M.M.

1995-01-01

Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

Plant life management strategies for pressurized heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

Using MCNP for in-core instrument calibration in CANDU

Energy Technology Data Exchange (ETDEWEB)

Taylor, D.C. [Point Lepreau Generating Station, NB Power, Lepreau, New Brunswick (Canada); Anghel, V.N.P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2002-07-01

The calibration of in-core instruments is important for safe and economical CANDU operation. However, in-core detectors are not normally suited to bench calibration procedures. This paper describes the use and validation of detailed neutron transport calculations for the purpose of calibrating the response of in-core neutron flux detectors. The Monte-Carlo transport code, MCNP, was used to model the thermal neutron flux distribution in the region around self-powered in-core flux detectors (ICFDs), and in the vicinity of the calandria edge. The ICFD model was used to evaluate the reduction in signal of a given detector (the 'detector shading factor') due to neutron absorption in surrounding materials, detectors, and lead-cables. The calandria edge model was used to infer the accuracy of the calandria edge position from flux scans performed by AECL's traveling flux detector (TFD) system. The MCNP results were checked against experimental results on ICFDs, and also against shading factors computed by other means. The use of improved in-core detector calibration factors obtained by this new methodology will improve the accuracy of spatial flux control performance in CANDU-6 reactors. The accurate determination of TFD based calandria edge position is useful in the quantitative measurement of changes in in-core component dimensions and position due to aging, such as pressure tube sag. (author)

Welding with the TIG automatic process of the end fittings for the execution of the Embalse nuclear power plant fuel channel rechange; Soldadura con proceso TIG automatico de los accesorios extremos (end fitting) para la ejecucion de un recambio de canal de combustible en el reactor de la Central Nuclear Embalse

Energy Technology Data Exchange (ETDEWEB)

Suarez, P O

1991-12-31

The present work describes the methodology for the cutting of the existing welding and subsequent welding applied by the TIG process of the coupling composed by the shroud ring and the end fitting ring from one of Embalse nuclear power plant`s fuel channels. The replacement will be previously determined by the SLAR-ETTE mechanism where a displacement operated among the Gartner Spring rings, the pressure tubes are separated from the Calandria tubes. The welding to be carried out has the function of stamping the CO{sub 2} annular gas (thermal insulator) circulating between the pressure tube and the Calandria one during the functioning of the plant. (Author). [Espanol] El presente trabajo describe la metodologia de corte de la soldadura existente y la posterior soldadura aplicada mediante proceso TIG de la junta compuesta por el aro de fuelle y el anillo del `end fitting`, de uno de los canales de combustibles del reactor de la Central Nuclear Embalse. El reemplazo, se determinara previamente mediante el mecanismo de SLAR-ETTE con lo cual se observara el desplazamiento operado entre los anillos garten spring que separan los tubos de presion de los tubos de calandria. La soldadura a efectuar cumple la funcion de sellar el gas anular CO{sub 2} (aislante termico) circulante entre el tubo de presion y el tubo de calandria durante el funcionamiento de la planta. (Autor).

Moderator circulation in CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Hussein, M.A.

1989-01-01

A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

Analysis of three idealized reactor configurations: plate, pin, and homogeneous

International Nuclear Information System (INIS)

McKnight, R.D.

1983-01-01

Detailed Monte Carlo calculations have been performed for three distinct configurations of an idealized fast critical assembly. This idealized assembly was based on the LMFBR benchmark critical assembly ZPR-6/7. In the first configuration, the entire core was loaded with the plate unit cell of ZPR-6/7. In the second configuration, the entire core was loaded with the ZPR sodium-filled pin calandria. The actual ZPR pin calandria are loaded with mixed (U,Pu) oxide pins which closely match the composition of the ZPR-6/7 plate unit cell. For the present study, slight adjustments were made in the atom concentrations and the length of the pin calandria in order to make the core boundaries and average composition for the pin-cell configuration identical to those of the plate-cell configuration. In the third configuration, the core was homogeneous, again with identical core boundaries and average composition as the plate and pin configurations

Preliminary Analysis For Wolsong Par Effects Using ISACC Calculations

International Nuclear Information System (INIS)

Song, Yong Mann; Kim, Dong Ha

2012-01-01

In the paper, hydrogen control effects using PARs only are analyzed for severe SBO station blackout (SBO) sequences beyond the design basis accidents in WS-1 which are of CANDU6 type reactor. As a computational tool, the latest version of ISAAC4.3 (Integrated Severe Accident Analysis Code for CANDU), which is a fully integrated and lumped severe accident computer code, is used to simulate hydrogen generation and transport inside the reactor building (R/B) before its failure. For the performance of hydrogen removal, the depletion rate equation of K-PAR developed in Korea is applied. In a CANDU reactor, three areas are identified as sources of hydrogen under severe accidents: fuel-coolant interactions in intact channels, suspended fuel or debris interactions in-calandria tank and debris interactions in-calandria vault. The first two origins provide source for the late ('late' terminology is used because it takes more than one day before calandria tank failure) potential hydrogen combustion before calandria tank failure and all the three origins would provide source for the very late potential hydrogen combustion occurring at or after calaria tank failure. If the hydrogen mitigation system fails, the AICC (adiabatic isochoric complete combustion) burning of highly flammable hydrogen may cause Wolsong R/B failure. So hydrogen induced failure possibility is evaluated, using preliminary ISAAC calculations, under several SBO conditions with and without PAR for both late and very late accident periods

Remote ultrasonic characterisation of an irradiated pressure tube from RAPS-II

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Muralidhar, S; Raut, S D; Ouseph, P M; Ghosh, J K; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Rajasthan Atomic Power Station Unit-2 (RAPS-2) has reached a stage of operation where the contacting pressure tubes are suspect to failure as a result of irradiation creep and displacement of the garter springs, the hot pressure tube coming in contact with the cold calandria tube. To study and assess the safety of these pressure tubes, two channels believed to be in contact with the calandria tubes, have been removed from the reactor for detailed full length post irradiation examination. Some of the test results are presented. 2 refs., 3 figs., 1 tab.

Assessment of ACR moderator circulation design using CFD

International Nuclear Information System (INIS)

Bunama, R.; Carlucci, L.N.; Waddington, G.M.

2004-01-01

Assessment of the thermalhydraulic performance of the moderator circulation system for the Advanced CANDU Reactor (ACR) was carried out using the specialized Computational Fluid Dynamics (CFD) code MODTURC C LAS V2.9 IST. The assessment included modeling the moderator circulation inside the calandria vessel under nominal and isothermal flow conditions. The modeling results show that the moderator flow through the core is relatively uniform and mostly upward. The moderator temperature distribution is nearly stratified and increases monotonically from the bottom to the top of the calandria vessel. (author)

Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

1999-10-01

In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

International Nuclear Information System (INIS)

Khartabil, H.F.

2000-01-01

A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

Study of sources, dose contribution and control measures of Argon-41 at Kaiga Generating Station

International Nuclear Information System (INIS)

Venkata Ramana, K.; Shrikrishna, U.V.; Manojkumar, M.; Ramesh, R.; Madhan, V.; Varadhan, R.S.

2001-01-01

Air is used as a medium for cooling calandria vault and thermal shield systems in the earlier Pressurised Heavy Water Reactors (Rajasthan Atomic Power Station and Madras Atomic Power Station) in India. This leads to production of significant quantity of 41 Ar in calandria vault and thermal shield cooling systems due to neutron activation of 40 Ar present in air (∼1% v/v). The presence of 41 Ar in reactor building contributes significant external doses to plant personnel during reactor operation and the release of this radionuclide to the environment result in dose to the public in the vicinity of the plants. An attempt is made to eliminate Argon-41 production in Indian standard Pressurised Heavy Water Reactors (Narora Atomic Power Station, Kakrapar Atomic Power Station, Kaiga Generating Station -1 and 2 and Rajasthan Atomic Power Station-3 and 4), by filling the calandria vault with demineralized water and providing a separate Annulus Gas Monitoring System (AGMS) for detecting leaks from calandria tube or pressure tube using Carbon dioxide as a medium. However, 41 Ar is produced in the Annulus Gas Monitoring System, Primary Heat Transport cover gas system and moderator cover gas system due to ingress of air into the systems during operational transients or due to trace quantity of air present as an impurity in the gases used for the above systems. A study was conducted to identify and quantify the sources of 41 Ar in the work areas. This report brings out the sources of 41 Ar, reasons for 41 Ar production and the results of the measures incorporated to reduce the presence of 41 Ar in the above systems. (author)

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

Pressure tube reactor

International Nuclear Information System (INIS)

Kanazawa, Nobuhiro; Kaneto, Kunikazu.

1979-01-01

Purpose: To attain uniform fluid poison distribution in a calandria tank by downwardly projecting, at an equal distance to the reactor core, a spacer wall from the periphery of an anti-vibration plate in the vicinity of a heavy water flow passage in the periphery of the anti-vibration plate, thereby decrease the amount of heavy water flowing into the heavy water flow passage. Constitution: A projecting wall concentrical with a calandria tank is suspended vertically from the boundary side at the peripheral portion of an anti-vibration plate to a water heavy flow passage in the periphery of the anti-vibration plate. The projecting wall has such a vertical length as about equal to the width of the heavy water flow passage, prevents heavy water flowing through apertures of a control rod guide tube from entering into the heavy water passage and increases the ratio of heavy water that flows through the heavy water flow passage in the anti-vibration plate. Consequently, if the liquid poison density in heavy water is varied, the ununiform poison density in the calandria tank can be prevented. (Seki, T.)

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

Full text: In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

DELOCA, a code for simulation of CANDU fuel channel in thermal transients

International Nuclear Information System (INIS)

Mihalache, M.; Florea, Silviu; Ionescu, V.; Pavelescu, M.

2005-01-01

In certain LOCA scenarios into the CANDU fuel channel, the ballooning of the pressure tube and the contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator arises through the contact area. If the temperature of channel walls increases, the contact area is drying, the heat transfer becomes inefficiently and the fuel channel could lose its integrity. DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after the contact between the two tubes. The code contains a few models: the creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This code was systematically verified by Contact1 and Cathena codes. This paper presents the results obtained at different temperature increasing rates. In addition, the contact moment for a RIH 5% postulated accident was calculated. The Cathena thermo-hydraulic code provided the input data. (authors)

Welding with the TIG automatic process of the end fittings for the execution of the Embalse nuclear power plant fuel channel rechange

International Nuclear Information System (INIS)

Suarez, P.O.

1990-01-01

The present work describes the methodology for the cutting of the existing welding and subsequent welding applied by the TIG process of the coupling composed by the shroud ring and the end fitting ring from one of Embalse nuclear power plant's fuel channels. The replacement will be previously determined by the SLAR-ETTE mechanism where a displacement operated among the Gartner Spring rings, the pressure tubes are separated from the Calandria tubes. The welding to be carried out has the function of stamping the CO 2 annular gas (thermal insulator) circulating between the pressure tube and the Calandria one during the functioning of the plant. (Author) [es

CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

International Nuclear Information System (INIS)

Chang, Se Myong; Park, A. Y.; Kim, Hyoung Tae

2011-01-01

The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

Validation of severe accident management guidance for the wolsong plants

International Nuclear Information System (INIS)

Park, S. Y.; Jin, Y. H.; Kim, S. D.; Song, Y. M.

2006-01-01

Full text: Full text: The severe accident management(SAM) guidance has been developed for the Wolsong nuclear power plants in Korea. The Wolsong plants are 700MWe CANDU-type reactors with heavy water as the primary coolant, natural uranium-fueled pressurized, horizontal tubes, surrounded by heavy water moderator inside a horizontal calandria vessel. The guidance includes six individual accident management strategies: (1) injection into primary heat transport system (2) injection into calandria vessel (3) injection into calandria vault (4) reduction of fission product release (5) control of reactor building condition (6) reduction of reactor building hydrogen. The paper provides the approaches to validate the SAM guidance. The validation includes the evaluation of:(l) effectiveness of accident management strategies, (2) performance of mitigation systems or components, (3) calculation aids, (4) strategy control diagram, and (5) interface with emergency operation procedure and with radiation emergency plan. Several severe accident sequences with high probability is selected from the plant specific level 2 probabilistic safety analysis results for the validation of SAM guidance. Afterward, thermal hydraulic and severe accident phenomenological analyses is performed using ISAAC(Integrated Severe Accident Analysis Code for CANDU Plant) computer program. Furthermore, the experiences obtained from a table-top-drill is also discussed

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

2004-02-01

To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

Development of channel inspection and gauging apparatus for 235 MWe PHWRs

International Nuclear Information System (INIS)

Parulkar, S.K.; Taneja, R.; Taliyan, S.S.; Singh, Manjit; Govindarajan, G.

1992-01-01

Channel inspection and gauging apparatus is being developed to enable in-service channel inspection and gauging. Phase I apparatus to measure annular gap between pressure tube and calandria tube in a dry channel has been developed. The apparatus consists of a gauging head and a drive mechanism. The gauging head utilities an eddy current probe to measure the annular gap between pressure tube and calandria tube and an ultrasonic sensor to measure the wall thickness of the pressure tube. The output signal of the eddy current probe needs to be corrected for the effect of pressure tube wall thickness variation. This paper gives the details of the above apparatus. The results of calibration tests at mock-up station are presented. The paper outlines the program for the phase-wise development of Channel Inspection and Gauging Apparatus for use in heavy water filled channels without their isolation from PHT and draining. The final apparatus will have the facilities for ultrasonic flaw detection, ultrasonic gauging to measure pressure tube diameter and wall thickness, an inclinometer to measure slope and sag of pressure tube and eddy current probe for the measurement of annular gap between pressure tube and calandria tube. (author). 6 figs

Reactor refurbishment options for a changing climate

Energy Technology Data Exchange (ETDEWEB)

McNeish, D. [Bruce Power, Tiverton, Ontario (Canada)

2012-07-01

As the industry looks ahead to another generation of reactor refurbishment, it is acknowledged that the traditional way of Retubing a reactor is a daunting prospect for our investors and stakeholders. Innovations are required to mitigate the long downtime and large one-time investment associated with previous reactor refurbishments. These can take the shape of improvements to the Retube processes or by fundamentally changing the approach, e.g., calandria/shield tank replacement or partial Retubes. This session presents technical challenges that utilities need help resolving to arrive at a more attractive reactor refurbishment model. This includes issues related to calandria vessel fitness-for-service, the fuel channel replacement process, the feeder replacement process, life extension of fuel channels and feeders and complexities involving interfacing systems. (author)

Reactor refurbishment options for a changing climate

International Nuclear Information System (INIS)

McNeish, D.

2012-01-01

As the industry looks ahead to another generation of reactor refurbishment, it is acknowledged that the traditional way of Retubing a reactor is a daunting prospect for our investors and stakeholders. Innovations are required to mitigate the long downtime and large one-time investment associated with previous reactor refurbishments. These can take the shape of improvements to the Retube processes or by fundamentally changing the approach, e.g., calandria/shield tank replacement or partial Retubes. This session presents technical challenges that utilities need help resolving to arrive at a more attractive reactor refurbishment model. This includes issues related to calandria vessel fitness-for-service, the fuel channel replacement process, the feeder replacement process, life extension of fuel channels and feeders and complexities involving interfacing systems. (author)

Thermal hydraulic simulation of moderator heat exchanger

International Nuclear Information System (INIS)

Anil Lal, S.; Rajakumar, A.; Vaidyanathan, G.; Srinivasan, R.; Chetal, S.C.

1993-01-01

Pressurized heavy water reactors form the majority in the first stage of India's nuclear power programme. Heavy water is both moderator and primary coolant. The heat generated in the moderator due to neutron moderation and capture has to be removed in moderator heat exchangers. It has been desired to improve the performance characteristics of moderator heat exchangers, whereby moderator would enter the calandria vessel at a low temperature and would enable higher power of operation for the same limiting temperature of moderator in the calandria. Results of studies carried out using a three dimensional computer code for various operating options are given. Using these velocities the heat exchangers have been analysed for flow induced vibrations. 7 refs., 6 figs., 6 tabs

Garter spring location of pressure tube for PHWR using eddy current testing methods

International Nuclear Information System (INIS)

Lee, Y. S.; Yang, D. J.; Jeong, H. K.

2001-01-01

There are garter springs between pressure tube and calandria tube for PHWR. If the space of these garter springs become to be changed, the sagging of tube is caused and the contact between the pressure tube and calandria tube will cause the tube to be failed. AECL has applied the eddy current testing methods using send-receive type probe for this purpose, but this study apply eddy current testing methods using bobbin differential type probe to detection of garter spring location. And we did the computer simulation using VIC-3D code and compared it with experiments results for inspection 1 ∼ 11kHz. The results was that the garter spring signal was successfully detected for every frequency, and 5 kHz was best

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

2011-01-01

In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

Thermal gradients caused by the CANDU moderator circulation

International Nuclear Information System (INIS)

Mohindra, V.K.; Vartolomei, M.A.; Scharfenberg, R.

2008-01-01

The heavy water moderator circulation system of a CANDU reactor, maintains calandria moderator temperature at power-dependent design values. The temperature differentials between the moderator and the cooler heavy water entering the calandria generate thermal gradients in the reflector and moderator. The resultant small changes in thermal neutron population are detected by the out-of-core ion chambers as small, continuous fluctuations of the Log Rate signals. The impact of the thermal gradients on the frequency of the High Log Rate fluctuations and their amplitude is relatively more pronounced for Bruce A as compared to Bruce B reactors. The root cause of the Log Rate fluctuations was investigated using Bruce Power operating plant information data and the results of the investigation support the interpretation based on the thermal gradient phenomenon. (author)

Cernavoda CANDU severe accident evaluation

International Nuclear Information System (INIS)

Negut, G.; Marin, A.

1997-01-01

The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

Method of and apparatus for use in joining tubular components and tube assemblies made thereby

International Nuclear Information System (INIS)

Percival, S.R.

1979-01-01

A method of joining difficult to weld materials involves the forming of a rolled joint. A particular application is the joining of zirconium alloy calandria tubes to stainless steel tube-plates in a SGHWR. (UK)

Manufacture of components for Canadian reactor programs

International Nuclear Information System (INIS)

Perry, L.P.

Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

Coolability of severely degraded CANDU cores. Revised

International Nuclear Information System (INIS)

Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

1996-01-01

Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

Safety benefits from CANDU reactor replacement - a case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

International Nuclear Information System (INIS)

Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

2011-01-01

In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

Safety benefits from CANDU reactor replacement. A case study

International Nuclear Information System (INIS)

Mottram, R.; Millard, J.W.F.; Purdy, P.

2011-01-01

Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)

Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

Consequences of pressure tube rupture on in-core components

International Nuclear Information System (INIS)

Hill, P.G.; Hauptmann, E.G.; Lee, V.

1982-12-01

An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

Development and validation of the 3-D CFD model for CANDU-6 moderator temperature predictions

International Nuclear Information System (INIS)

Yoon, Churl; Rhee, Bo Wook; Min, Byung Joo

2003-03-01

A computational fluid dynamics model for predicting the moderator circulation inside the CANada Deuterium Uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the Calandria tubes. The buoyancy effect induced by internal heating is accounted for by Boussinesq approximation. The standard κ-ε turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the Calandria tubes in the core region is simplified to porous media, in which an-isotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The CFD model has been successfully verified and validated against experimental data obtained in the Stern Laboratories Inc. (SLI) in Hamilton, Ontario

Analytical model for performance verification of liquid poison injection system of a nuclear reactor

International Nuclear Information System (INIS)

Kansal, Anuj Kumar; Maheshwari, Naresh Kumar; Vijayan, Pallippattu Krishnan

2014-01-01

Highlights: • One-dimensional modelling of shut down system-2. • Semi-empirical correlation poison jet progression. • Validation of code. - Abstract: Shut down system-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1D) hydraulic code, COPJET is developed, to predict the performance of system by predicting progression of poison jet with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for advanced vertical pressure type reactor

Ion chamber repairs in Bruce A

International Nuclear Information System (INIS)

Millard, J.; Edwards, T.; Kerker, J.; Pletch, R.; Edwards, T.

2012-01-01

This paper discusses identification and successful remediation of leakage of shield tank water on vertical and horizontal Ion Chambers in Bruce A. In doing so, it discusses real events moving from the initial investigation to understand the problem, through looking at options for solutions, and moving to site work and actual resolution.. In multiunit 900 MW class CANDU® reactors, the calandria vessel is suspended within a larger shield tank. Due to temperature changes or changes in moderator fluid levels in the calandria, the calandria can move relative to the shield tank and its reactivity deck. Thimbles which contain the reactivity sensors and controls connect the two vessels and allow the reactivity drives and controls connections to be placed on the deck structure on the top of the reactor assembly for RRS and SDS1 and horizontally for SDS2. These thimbles have expansion joints with metal bellows where they meet the deck structure or shield tank walls. The deck structure lies on a vault containment boundary. The horizontal ion chambers are not in the containment boundary as they connect the outside of the calandria and shield tank around mid plane in the reactor vault, but due to geometry difference provides a more challenging work environment. Bruce had a beetle alarm (1-63851-MIA2-ME30 in alarm state (vertical IC housing)) at the start of April 2012 on Unit 1 channel F vertical Ion chamber expansion joint at the deck connection. This occurred after the moderator levels had been raised after the several years long refurbishment outage and the expansion joint had a significant travel. The investigation showed shield tank water in the collection chamber at the beetle. In addition, Channel J of the horizontal ion chamber had a seized instrument, which on removal was found to relate to oxide build up as a result of minor water leakage into the site. Repairs in both cases were performed as part of the long Bruce 1 & 2 refurbishment outage to completely stop the

Analysis of severe core damage accident progression for the heavy water reactor

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

CATHENA Analysis Of Candu Advanced Passive Moderator Concept In Normal Operation Condition

International Nuclear Information System (INIS)

Alfa, Sudjatmi K

2001-01-01

In the CANDU - advanced passive moderator (APM) concept, the positive void reactivity is eliminated by reducing the density of the moderator. The simple model for the CANDU APM concept consists of the calandria, heat exchanger, pump, and a stabilizing tank, along with connecting piping. The calandria is divided into two parts, one part simulates the down area, while the other simulates up flow area. To demonstrate the thermalhydraulic behavior of the APM concept, Canadian algorithm for thermalhydraulic network analysis (CATHENA) code is used. The simulation for a pressure boundary condition of 300, 330 and 360 kPa and for water coolant mass flow rate boundary conditions of 2000 and 3000 kg/s respectively have been studied. Preliminary results show that there is boiling in the core, with vapor condensing in the heat exchanger. It is important to note, that the solution had not reached steady state when the boiling occurred

Coolability of severely degraded CANDU cores

International Nuclear Information System (INIS)

Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Mijhawan, S.

1995-07-01

Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

Fission-product releases from a PHWR terminal debris bed

Energy Technology Data Exchange (ETDEWEB)

Brown, M.J.; Bailey, D.G., E-mail: morgan.brown@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

MacKinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

1997-01-01

During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system conditions investigated are of a reactor in a GSS, with coolant in the primary heat transport system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The

Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

Science.gov (United States)

Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

2018-04-01

Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

Supplementary shutdown system of 220 MWe standard PHWR in India

International Nuclear Information System (INIS)

Muktibodh, U.C.

1997-01-01

The design objective of the shutdown system is to make the reactor subcritical and hold it in that state for an extended period of time. This objective must be realised under all anticipated operational occurrences and postulated abnormal conditions even during most reactive state of the core. PHWR design criteria for shutdown stipulates requirement of two independent diverse and fast acting shutdown systems, either of which acting alone should meet the above objectives. This requirement would normally call for a large number of reactivity mechanism penetrations into the calandria. From the point of view of space availability at the reactivity mechanism area on top of calandria, for the relatively small core of 220 MWe PHWRs, and ease of maintenance realisation of the total worth by either of the shutdown systems acting alone was difficult. To overcome this engineering constraint and at the same time to satisfy the design criteria, a unique approach to meet the reactivity demands for shutdown was adopted. The reactivity requirements of the shutdown consists of fast and slow reactivity changes. For the shutdown system of 220 MWe PHWRs, the approach of realizing fast reactivity changes with dual redundant, diverse, fast acting shutdown systems aided by a slow acting shutdown system to counter delayed reactivity changes was conceived. The supplementary slow acting shutdown system is called upon to act after actuation of either of the two redundant fast acting systems and is referred to as Liquid Poison Injection System (LPIS). The system adds bulk amount of neutron poison (boric acid), equivalent to 45 mk, directly into the moderator through two nozzles in calandria using pneumatic pressure. This paper describes the design of LPIS as envisaged for the standardised 220 MWe PHWRs. (author)

Station black out analysis for CANDU 6 plant

International Nuclear Information System (INIS)

Baburajan, P.K.; Rao, R.S.; Gupta, S.K.

2011-01-01

As part of International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), 'Benchmarking severe accident computer codes for pressurised heavy water reactor applications', thermal hydraulic analysis of severe accident station black out (SBO) is carried out for a generic CANDU 6 plant. The CRP is conducted in order to improve severe accident analysis capability for heavy water reactors (HWRs) through the benchmarking exercise. The plant simulation is carried out using RELAP5/Mod3.4 best estimate system thermal hydraulic code. The total thermal power of the plant is 2064 MW. There are 380 fuel channels in the core, 12 fuel bundles per channel and each bundle assembly has 37 fuel elements. The primary heat transport system (HTS) consists of two loops. Each loop consist of inlet and exit headers, feeder lines, fuel channels, hot leg and cold leg of steam generator, pumps, pump suction and discharge lines. Ninety five fuel channels in each pass of the loop are simulated as a single channel. The steam generator as the secondary side heat sink consists of annulus down-comer, riser, steam separator, steam drum, steam header and steam lines. Fuel channels (pressure tube) and calandria tube are simulated using SCDAPSIM to study the severe accident code behaviour. The SBO transient is initiated after obtaining the steady state conditions. Present analysis is carried out till the pressure tube failure. Analysis results show that the secondary inventory is lost in about 6500 seconds of the transient. The primary inventory is lost in 10370 seconds of the transient and subsequently the pressure tube failure is predicted as the tube wall temperature exceeded 900 K. Further analysis is to be carried out by incorporating changes in the calandria model and including the modeling of calandria vault and containment. (author)

Report of the Federal Ministry for the Environment, Nature Conservation, Buildings and Nuclear Safety (BMUB) on the topical peer review aging management in nuclear power plants and research reactors; Bericht des Bundesministeriums fuer Umwelt, Naturschutz, Bau und Reaktorsicherheit (BMUB) zum Topical Peer Review Alterungsmanagement in Kernkraftwerken und Forschungsreaktoren

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-12-28

The report of the Federal Environmental Ministry (BMUB) on the topical peer review aging management in nuclear power plants and research reactors covers the following issues: comprehensive requirements for aging management and its implementation, electric cables, non accessible pipes, reactor pressure vessel, calandria/pressure tubes (CANDU), concrete containment, pre-stressed concrete reactor pressure vessel (AGR).

Report of the Federal Ministry for the Environment, Nature Conservation, Buildings and Nuclear Safety (BMUB) on the topical peer review aging management in nuclear power plants and research reactors

International Nuclear Information System (INIS)

2017-01-01

The report of the Federal Environmental Ministry (BMUB) on the topical peer review aging management in nuclear power plants and research reactors covers the following issues: comprehensive requirements for aging management and its implementation, electric cables, non accessible pipes, reactor pressure vessel, calandria/pressure tubes (CANDU), concrete containment, pre-stressed concrete reactor pressure vessel (AGR).

Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

Energy Technology Data Exchange (ETDEWEB)

Soliman, S A; Lee, T; Ibrahim, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs.

Ageing of coolant channels in nuclear reactors (PHWRs)

International Nuclear Information System (INIS)

Mitra, T.L.; Chowdhury, M.K.; Gupta, R.K.; Pandarinathan, P.R.; Seth, V.K.

1994-01-01

In PHWRs, ageing of various components takes place due to factors like fast neutron flux, temperature, stress, environment etc. In coolant channel, the most severely affected component due to ageing is pressure tube, though other components like end fitting, calandria tube, garter spring spacer also show ageing to a limited extent. Ageing effects in pressure tube are seen in the form of diametral and axial creep, corrosion, delayed hydrogen cracking and irradiation hardening. In calandria tube and garter spring spacer, creep and hardening are seen though these are not of concern in PHWRs. In end fitting, irradiation embrittlement and abrasion of sealing faces are the areas of concern. Ageing process in these components are the areas of concern. Ageing process in these components are effectively retarded by taking measures like selection of proper material, manufacturing process, control of environmental chemistry, and design modifications. Experience and information gained in various Indian and foreign reactors have been used to improve upon the design in 220 MWe reactors and have formed the basis of design for 500 MWe reactors. (author). 3 refs., 5 figs

The simulation of CANDU fuel channel behavior in thermal transient conditions

International Nuclear Information System (INIS)

Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

2005-01-01

In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

International Nuclear Information System (INIS)

Soliman, S.A.; Lee, T.; Ibrahim, A.M.; Hodgson, S.

1995-01-01

This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs

Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment

International Nuclear Information System (INIS)

Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las

1965-01-01

This exponential experiment required 74 units (37 loaded with UO 2 and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

Evaluation of A-1 reactor heavy-water calandria specimens

International Nuclear Information System (INIS)

Brumovsky, M.

1976-01-01

Container chains with surveillance specimens were placed in two special channels of the core peripheral part to test changes in mechanical properties due to reactor operation of caisson tube material. The specimens were made from the caisson tube material and placed by eight pieces on the outer surface of the containers. The first removed specimens were tested for corrosion losses, tensile strength, and fractured surfaces were then assessed. The changes in strength properties were found to be similar in both base material and welded joints. The corrosion film on surveillance specimens did not practically affect strength properties nor ductility. It was found that the Al-Mg-Si alloy used for the heavy water vessel caisson tubes following stabilization annealing was fully stable at operating temperatures of up to 100 degC. Slio.ht changes in properties can be attributed to the effect of a high neutron dose. Thus, the high radiation and temperature stability of the alloy was confirmed. (O.K.)

Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

Energy Technology Data Exchange (ETDEWEB)

Diaz Beltran, A; Jaraiz Franco, E; Rivas Diaz, M. de las

1965-07-01

This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

3-D CFD analysis of the CANDU-6 moderator circulation under normal operating conditions

International Nuclear Information System (INIS)

Yoon, Churl; Rhee, Bo Wook; Min, Byung Joo

2004-01-01

A computational fluid dynamics model for predicting moderator circulation inside the Canada Deuterium Uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard k-ε turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are 'mixed-type', as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is 82.9 deg. C at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is 24.8 .deg. C

Fuel Management Study for a CANDU reactor Using New Physics Codes Suite

International Nuclear Information System (INIS)

Kim, Won Young; Kim, Bong Ghi; Park, Joo Hwan

2008-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. The primary reactivity control in a CANDU reactor is the on-power refueling on a daily basis and an additional reactivity control is provided through an individual reactivity device movement, which includes 21 adjusters, 6 liquid zone controllers, 4 mechanical control absorbers and 2 shutdown systems. The refueling in CANDU is carried out on power and this makes the in-core fuel management different from that in a reactor refueled during shutdowns. The objective of a fuel management is to determine a fuel loading and fuel replacement procedure which will result in a minimum total unit energy cost in a safe and reliable operation. In this article, the in-core fuel management for the CANDU reactor was studied by using the new physics code suite of WIMS-IST/DRAGON-IST/RFSP-IST with the model of Wolsong-1 NPP

Eddy current detection of spacers in the fuel channels of CANDU nuclear reactors

International Nuclear Information System (INIS)

Krause, T.W.; Schankula, J.; Sullivan, S.P.

2002-01-01

Garter Spring (GS) spacers in the fuel channels of CANDU nuclear reactors maintain separation between the hot pressure tube and surrounding moderator cooled calandria tube. Eddy current detection of the four GSs provides assurance that spacers are at or close to design positions and are performing their intended function of maintaining a non-zero gap between pressure tube and calandria tube. Pressure tube constrictions, resulting from relatively less diametral creep at end-of-fuel bundle locations, also produce large eddy current signals. Large constrictions, present in higher service pressure tubes, can produce signals that are 10 times larger than GS signals, reducing GS detectability to 30% in standard GS-detect probes. The introduction of field-focussing elements into the design of the standard GS detection eddy current probe has been used to recover the detectability of GS spacers by increasing the signal amplitude obtained from GSs relative to that from constrictions by a factor of 10. The work presented here compares laboratory, modelling and in-reactor measurements of GS and constriction signals obtained from the standard probe with that obtained from field-focussed eddy current probe designs. (author)

Current safety issues of CANDU licensing

International Nuclear Information System (INIS)

Lee, Y.; Natalizio, A.

1994-01-01

As requested by Korea Institute of Nuclear Safety(KINS), the status of five generic licensing issues has been examined and their potential impact on a new plant that would be constructed in Canada has been evaluated. The results and conclusions of this evaluation are summarized as follows: steam explosion in calandria, hydrogen explosion in containment, use of PSA in reactor licensing, human factors, safety critical software

Humid scraping method to obtain samples for the analysis of D2 incorporated in the pressure tubes of Embalse Nuclear Power Plant

International Nuclear Information System (INIS)

Binetti, Edgardo O.; Cerutti, Carlos R.

1999-01-01

From ten fuel channels of the CNE reactor four samples of each channel were taken by means of the Humid Scraping method in order to evaluate the equivalent hydrogen content by incorporating deuterium in the pressure tubes. With these data, it is possible to make a list of priorities of channels for future replacement of spacer rings between pressure and calandria tubes, using Slarette equipment. (author)

Jet flow analysis of liquid poison injection in a CANDU reactor using source term

International Nuclear Information System (INIS)

Chae, Kyung Myung; Choi, Hang Bok; Rhee, Bo Wook

2001-01-01

For the performance analysis of Canadian deuterium uranium (CANDU) reactor shutdown system number 2 (SDS2), a computational fluid dynamics model of poison jet flow has been developed to estimate the flow field and poison concentration formed inside the CANDU reactor calandria. As the ratio of calandria shell radius over injection nozzle hole diameter is so large (1055), it is impractical to develop a full-size model encompassing the whole calandria shell. In order to reduce the model to a manageable size, a quarter of one-pitch length segment of the shell was modeled using symmetric nature of the jet; and the injected jet was treated as a source term to avoid the modeling difficulty caused by the big difference of the hole sizes. For the analysis of an actual CANDU-6 SDS2 poison injection, the grid structure was determined based on the results of two-dimensional real- and source-jet simulations. The maximum injection velocity of the liquid poison is 27.8 m/s and the mass fraction of the poison is 8000 ppm (mg/kg). The simulation results have shown well-established jet flow field. In general, the jet develops narrowly at first but stretches rapidly. Then, the flow recirculates a little in r-x plane, while it recirculates largely in r-θ plane. As the time goes on, the adjacent jets contact each other and form a wavy front such that the whole jet develops in a plate form. his study has shown that the source term model can be effectively used for the analysis of the poison injection and the simulation result of the CANDU reactor is consistent with the model currently being used for the safety analysis. In the future, it is strongly recommended to analyze the transient (from helium tank to injection nozzle hole) of the poison injection by applying Bernoulli equation with real boundary conditions

PIV Measurement of Isothermal Flow in the Moderator Circulation Test (MCT) Facility

Energy Technology Data Exchange (ETDEWEB)

Im, Sunghyuk; Sung, Hyung Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

One of the important design features of a CANDU reactor (a pressurize heavy water reactor) is the use of moderator as a heat sink during some postulated accidents such as a large break Loss Of Coolant Accident (LOCA). If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as the fuel channel contact boiling experiments. The difference between available subcooling and required subcooling is called subcooling margins. The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the local temperature in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. In the present work the test vessel is equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV is performed under the iso-thermal test conditions. The 2D velocity is measured on the cross-sectional plane normal to the axial direction of the tank. The PIV measurement results could capture the same flow pattern as that expected in the CANDU6 calandria tank under momentum dominant flow condition, where the inlet jets penetrate to the top of the tank and produce a downward flow through the center of the tube columns towards the outlet nozzle and the flow fields are in symmetric distributions. The measurements of downward velocities are performed at different locations. The velocity is shown to be axially uniform. The velocity is rapidly decreased as the measurement location is far from the center of tank, since the downward flow is dominant along the center of the tube columns. More experimental works for the iso-thermal conditions as well as the heating conditions will be performed using PIV measurement in the

ACR-1000 design provisions for severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

Preliminary test results and CFD analysis for Moderator Circulation Test (MCT)

International Nuclear Information System (INIS)

Kim, Hyoung Tae

2015-01-01

Highlights: • Korea Atomic Energy Research Institute (KAERI) installed the Moderator Circulation Test (MCT) facility. • Velocity profiles for iso-thermal conditions are measured by the Particle Image Velocimetry (PIV). • The PIV measurement results can capture the same flow pattern as that expected in the CANDU6 calandria tank under a momentum dominant flow condition. • More experimental works for the iso-thermal conditions as well as the heating conditions will be performed. • The CFX model will be validated against the PIV measurement data in the future. - Abstract: The moderator flow circulation patterns in CANDU6 reactor are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. The code and its analytical models have therefore to be validated against experiments representative of reactor conditions. Korea Atomic Energy Research Institute (KAERI) installed the Moderator Circulation Test (MCT) facility to simulate the 3 dimensional moderator circulation phenomena in the calandria of CANDU6 reactor and develop the optical measurement system using the Particle Image Velocimetry (PIV). From the present work it is shown that the PIV measurement results can capture the same flow pattern as that expected in the CANDU6 calandria tank under a momentum dominant flow condition, where the inlet jets penetrate the top of the tank and produce a downward flow through the center of the tube columns toward the outlet nozzle, and the flow fields are in symmetric distributions. The measurements of the downward velocities are performed at different locations. The velocity is shown to be axially uniform. The velocity is rapidly decreased as the measurement location is far from the center of the tank, since the downward flow is dominant along the center of the tube columns

PIV Measurement of Isothermal Flow in the Moderator Circulation Test (MCT) Facility

International Nuclear Information System (INIS)

Im, Sunghyuk; Sung, Hyung Jin; Seo, Han; Bang, In Cheol; Kim, Hyoung Tae

2014-01-01

One of the important design features of a CANDU reactor (a pressurize heavy water reactor) is the use of moderator as a heat sink during some postulated accidents such as a large break Loss Of Coolant Accident (LOCA). If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as the fuel channel contact boiling experiments. The difference between available subcooling and required subcooling is called subcooling margins. The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the local temperature in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. In the present work the test vessel is equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV is performed under the iso-thermal test conditions. The 2D velocity is measured on the cross-sectional plane normal to the axial direction of the tank. The PIV measurement results could capture the same flow pattern as that expected in the CANDU6 calandria tank under momentum dominant flow condition, where the inlet jets penetrate to the top of the tank and produce a downward flow through the center of the tube columns towards the outlet nozzle and the flow fields are in symmetric distributions. The measurements of downward velocities are performed at different locations. The velocity is shown to be axially uniform. The velocity is rapidly decreased as the measurement location is far from the center of tank, since the downward flow is dominant along the center of the tube columns. More experimental works for the iso-thermal conditions as well as the heating conditions will be performed using PIV measurement in the

Advances in fuel channel technology for CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.

1994-05-01

The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

1978 annual report

International Nuclear Information System (INIS)

1978-06-01

In fiscal 1978 efforts continued to be made to increase New Brunswick's energy independence through research into and development of indigenous power sources including coal, peat, tidal, and hydroelectric resources. At Point Lepreau nuclear generating station the reactor building was completed, the concrete vault was prepared, and the calandria was installed. Work continued within the reactor building. Excavation of the cooling water tunnels and riser shafts was completed. (LL)

An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

International Nuclear Information System (INIS)

Gillespie, G.E.

An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

An emergency water injection system (EWIS) for future CANDU reactors

International Nuclear Information System (INIS)

Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

2000-01-01

This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m 2 : the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

MODTURCCLAS analysis of moderator poison/coolant mixing in the calandria due to a pressure tube/calandria tube guillotine rupture during an overpoisoned guaranteed shutdown state

International Nuclear Information System (INIS)

Mackinnon, J.C.; Szymanski, J.K.; Balog, G.

1996-01-01

This paper reports the results of a study to investigate moderator poison/coolant mixing due to a guillotine rupture of a fuel channel when the reactor is in an overpoisoned guaranteed shutdown state. The analysis, performed using MODTURC C LAS, allowed for study of the mixing characteristics and the spatial and temporal evolution of the concentration fields. Results for simulated breaks at three channel locations show that the poison in the vessel is quite well mixed throughout the transient, resulting in no extensive regions of low poison concentration. MODTURC C LAS calculations show that at all three break locations investigated, the displacement of poison from the vessel through the relief ducts is less than that calculated by both the simple uniform mixing model and piston mixing model. This result is expected to hold for all break locations in the core. (author)

Methodologies for assessment of the service life of pressure tubes in Indian PHWRs

International Nuclear Information System (INIS)

Sinha, R.K.; Sharma, A.; Madhusoodanan, K.; Sinha, S.K.; Malshe, U.D.

1997-01-01

For estimating safe service life of pressure tubes in Indian PHWRs, analytical methodologies have been developed to evaluate creep deformation, deuterium pick-up rate, blister growth at cold spot, and operating domain required for achieving leak-before-break. The paper provides an overview of these methodologies, and results of some studies carried out towards evolution of proposed fitness-for-service criteria for a pressure tube in contact with its calandria tube. (author)

Operating performance of CANDU pressure tubes

International Nuclear Information System (INIS)

Cheadle, B.A.; Price, E.G.

1989-04-01

The performance of Zircaloy-2 and Zr-2.5 Nb pressure tubes in CANDU reactors is reviewed. The accelerated hydriding of Zircaloy-2 in reducing water chemistries can lower the toughness of this material and it is essential that defect-initiating phenomena, such as hydride blister formation from pressure tube to calandria tube contact, be prevented. Zr-2.5 Nb pressure tubes are performing well with low rates of hydrogen pick-up and good retention of material properties

Fine element (F.E.) modelling of hydrogen migration and blister formation in PHWR coolant channels

International Nuclear Information System (INIS)

Prasad, P.S.; Dutta, B.K.; Sinha, R.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1995-01-01

The formation of a cold spot in pressure tube due to its contact with calandria tube of PHWR coolant results in the migration of Hydrogen in pressure tube towards contact zone from its surrounding material. A 3-D finite element code SPARSH is developed to model the hydrogen redistribution and consequent hydride blister formation due to thermal and Hydrogen concentration gradients. In the present paper, the details and performance of this code are presented. (author). 6 refs., 2 figs

Refinements to calandria tube - liquid injection nozzle (CT-LIN) contact assessments

International Nuclear Information System (INIS)

Sedran, P.J.

2012-01-01

In recent years, the issue of CT-LIN contact, which first gained attention in 1989, has been addressed through CT-LIN gap measurements, followed by analytical predictions of time-to-contact. CT-LIN time-to-contact predictions have been preformed independently by CPUS Limited for Point Lepreau and Gentilly-2 and by AECL Sheridan Park (now Candu Energy Inc.) for Bruce Power and Gentilly-2. Both companies used the CDEPTH code in combination with CT-LIN gap measurements. Subsequent to the assessments for Point Lepreau and Gentilly-2, a recommended approach for future assessments was presented at the 2008 CANDU maintenance conference. Since that time, a number of refinements to the overall strategy for predicting CT-LIN time-to-contact have been developed and are outlined in this paper. The refinements include: 1. The use of ultrasonic LIN elevation measurements to confirm LIN creep sag behaviour 2. The development of a non-linear empirical CT Creep Sag Model 3. The development of a rationale for discrepancies observed in repeated optical CT-LIN gap measurements and a discussion of alternative CT-LIN gap measurements With these refinements, more accurate CT-LIN time-to-contact predictions can be obtained. For stations that plan to refurbish by 210,000 EFPH, the improvement in time-to-contact predictions resulting from the fore mentioned refinements will not be of any real benefit.. However, for stations that are planning life extensions in order to operate beyond 210,000 EFPH, CT-LIN contact will be an issue. For these stations, improvements in CT-LIN contact time predictions would be beneficial. This paper presents a summary of the proposed refinements and demonstrates how they would impact CT-LIN time-to-contact predictions. (author)

Evolution of criteria for repair work on helium lines of Cirus reactor

International Nuclear Information System (INIS)

Mishra, Rajesh; Soni, R.S.; Kushwaha, H.S.

2006-05-01

The research reactor CIRUS uses light water as coolant and heavy water as moderator and is rated for a thermal power of 40 MW. This reactor has been in operation since 1960 and has undergone refurbishment work recently. In the CIRUS reactor, helium gas is utilised as the cover gas. The helium lines are connected with the tube sheet at the top of the calandria. There are eight such helium lines at the top of the calandria, out of which four are connected to one ring header, three to another ring header and the remaining one is single line. These helium gas lines have tongue and groove joints for connecting the stainless steel piping with the aluminium piping. With the prolonged operation of the plant, leakage was observed at these joints. As a part of reactor refurbishing work, these joints were required to be repaired. Since these joints are situated in an inaccessible area, the entire job was to be carried out remotely and therefore, a fail-safe scheme was to be evolved based on computer simulation and analytical work. The entire analysis work had many challenging aspects hence, utmost care was exercised while analytically formulating the scheme for the tightening of these flange joints by postulating the various possible scenarios and by maintaining the stress level within the limits, particularly at the fillet welds between the aluminium pipe and calandria tube sheet. Another challenging aspect of this job was to take care of various uncertainties regarding the prevailing status of the joints. This report highlights the methodology adopted to arrive at the optimum amount of tightening and sequence of tightening. This report also highlights how analytical simulation of actual site scenario was carried out based on site feedbacks at various stages of tightening operations and how strategies were formulated to overcome various challenges and also to take care of various uncertainties in the input information being reported by the site. The tightening work

Leak-before-break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-01-01

In the Canada deuterium uranium (CANDU) reactor, each of the ∼ 400 hot pressure tubes containing the fuel bundles and the pressurized heat transport water is surrounded and insulated from the cold moderator by a calandria tube. The pressure tubes are made from cold-worked Zr-2.5 Nb with a minimum wall thickness of 4.19 mm, and the calandria tubes are made from annealed Zircaloy-2 with a minimum wall thickness of 1.37 mm. The annulus between these two tubes contains an inert gas. Leak-before-break has developed into an operational tool in CANDU reactors to prevent unstable failure of pressure tubes. A procedure for leak detection and reactor response has been developed from the use of the annulus gas, whose dew point is measured to ascertain if leaks have crept into the annulus. The characteristics of the crack are used to establish the response time for leak detection. The reactor is required to be shut down before the length of the slowly growing crack has reached the critical stage. This critical crack length, determined using slit burst tests on tubes, is the crack length at which the crack growth becomes unstable. The most likely crack growth mechanism is delayed hydride cracking. This mechanism requires three conditions to occur simultaneously: the material must be sensitive to delayed hydride cracking; zirconium hydrides must be present in the material; and the tensile stress must be sufficiently great

Transient Moderator Simulation Using CFX10-CAMO, a CANDU Moderator Analysis Model Based on a Coupled Solver

International Nuclear Information System (INIS)

Yoon, Churl; Park, Joo Hwan

2007-01-01

When a PHT(Primary Heat Transfer) system fails to remove excess heat from fuel channels for some loss of coolant accidents(LOCA's) in CANDU NPP's, the fuel channel temperature could increase until the pressure tube strains (i.e., balloon or sag) to contact its surrounding Calandria tube.(PT/CT contact) Following a PT/CT contact, there is a spike in the heat flux to the moderator surrounding the Calandria tube, which may lead to a sustained CT dryout and also a failure of a fuel channel. The prevention of a CT dryout following a PT/CT contact depends on the local moderator subcooling. That is, fuel channel integrity depends on the capability of the moderator to act as the ultimate heat sink for some LOCA's in a CANDU reactor. In KAERI, Yoon et al. developed a CFD model for predicting a CANDU-6 moderator temperature on the basis of a commercial CFD code CFX-4(ANSYS Inc.). This analytic model has the strength of modelling the hydraulic resistances in the core region and accounting for a heat source term in the energy equations. But convergence difficulties and a slow computing speed are the limitations of this model, because the CFX-4 code adapts a segregated solver to resolve a moderator circulation including a strong coupled-effect. Compared to a segregated solver, a coupled-solver is highly efficient and robust especially for a flow with a strong interference between the variables such as combustion

Validation of a CFD analysis model for the calculation of CANDU6 moderator temperature distribution

International Nuclear Information System (INIS)

Yoon, Churl; Rhee, Bo Wook; Min, Byung Joo

2001-01-01

A validation of a 3D CFD model for predicting local subcooling of moderator in the vicinity of calandria tubes in a CANDU reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory (SPEL) in Ontario, Canada is used for the validation. Also a comparison is made between previous DFD analyses based on 2DMOTH and PHOENICS, and the current model analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard κ-ε turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used and buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is a buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than 2.0 .deg. C over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well

Validation of a CFD Analysis Model for Predicting CANDU-6 Moderator Temperature Against SPEL Experiments

International Nuclear Information System (INIS)

Churl Yoon; Bo Wook Rhee; Byung-Joo Min

2002-01-01

A validation of a 3D CFD model for predicting local subcooling of the moderator in the vicinity of calandria tubes in a CANDU-6 reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory (SPEL) in Ontario, Canada[1] is used for the validation. Also a comparison is made between previous CFD analyses based on 2DMOTH and PHOENICS, and the current analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard k-ε turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used. Buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is the buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than 2.0 deg. C over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well. (authors)

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

Assessments of long term mechanical behavior of CANDU fuel channel by means of PFEM analysis

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2005-01-01

Full text: Structural analysis with finite elements method is today a usual way to evaluate and predict the behavior of structural assemblies exposed to severe conditions, in order to ensure their safety end reliability. CANDU 600 fuel channel is an example in which long time irradiation with implicit consequences on material properties evolution interfere with the corrosion and thermal aggression. A high degree of uncertainty in the evolution of the material's properties must be considered. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods, in order to predict the structural components response. In INR (Institute of Nuclear Research) in the past years, a code for thermo-mechanical analysis of the fuel channel from CANDU 600 nuclear power plant of Cernavoda was developed using finite element methods (FEM). The CANTUP code evaluate the stress and strain state of the mechanical assembly of fuel channel, considered to be divided into pressure tube, calandria tube and four spacers, supposed to be equidistantly distributed along the pressure tube. The main achievement obtained with this code was the prediction of the long-term behavior of the sag of the pressure tube, by analysis in which the creep phenomenon and the contact between the spacers and calandria tube were considered. This reason has sustained the attempt to estimate the possibility to use this code in order to perform probabilistic evaluations. (authors)

Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

International Nuclear Information System (INIS)

Dwivedi, K.P.; Seth, V.K.

1994-01-01

A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

Strategies for accelerating the SLARette process

International Nuclear Information System (INIS)

Grewal, P.

1997-01-01

The SLARette (Spacer Location and Repositioning) process is continuing on several CANDU reactors, where loose fitting garter springs (spacers) were used, to prevent contact between the calandria tube and the pressure tube for the target life. With time, the sag in the fuel channel is increasing and consequently increasing the potential for contact between the pressure tube and the calandria tube. Also, due to increasing sag in the pressure tubes and increasing magnitude of the fuel channel constrictions on the eddy current detection system, the Spacer Location and Repositioning activities are becoming more time consuming and difficult. For CANDU owners, during the SLARette campaigns, station outage time is the most expensive item. Therefore, it is beneficial to complete the SLARette process as early as possible and as fast as possible. New SLARette strategies can substantially accelerate the overall SLARette process and thus minimize the outage time. There are several strategies to perform the SLARette process. These strategies include: using the SLARette Mark II Delivery System; using the SLARette Advanced Delivery System; implement creative fuel handling technique; operate from both sides of the reactor using Mark II Delivery Systems; operate both sides using Advanced Delivery Systems. Each strategy offers different benefits, rate of fuel channel processing (SLARette Activity), and schedule constraints. This paper provides the details of each strategy and compare them in terms of outage time, man-rem consumption, and constraints. (author)

Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

Energy Technology Data Exchange (ETDEWEB)

Anca Abati, R de

1966-07-01

It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs.

The development of a remote gauging and inspection capability for fuel channels in Candu reactors

International Nuclear Information System (INIS)

Dolbey, M.P.; Kupcis, O.A.

1979-01-01

Equipment under development for the inspection and gauging of pressure tubes in CANDU (Canadian Deuterium Uranium) type reactors is described. A brief overview of the mechanical scanning system is presented followed by a detailed description of the measurement and data processing systems for the gauging of diameter and wall thickness, volumetric inspection of the tube wall and gauging of the annular gap between the pressure tube and the calandria tube. Experience of testing ultrasonic transducers in very high (10 6 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

Fuel elements assembling for the DON project exponential experience

International Nuclear Information System (INIS)

Anca Abati, R. de

1966-01-01

It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs

Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1986-01-01

During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

Dynamic analysis of a reactor building on alluvial soil

International Nuclear Information System (INIS)

Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.

1977-01-01

The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular grid slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected structures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumptions required to be made in developing the mathematical model are briefly discussed in the paper. Transfer matrix technique has been used to determine the frequencies and mode shapes. The deformations due to bending, shear and effect of the rotary inertia have been included. Various alternatives of laterally interconnecting the internals and the shells have been examined and the best alternative from earthquake considerations has been obtained. In the study, the effect of internal structure flexibility and Calandria vault flexibility on the whole building have been studied. The resulting base raft motion and the structural timewise response of all floors have been determined for the design basis (safe shutdown) earthquake by mode superposition

Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

International Nuclear Information System (INIS)

Aydogdu, K.

1998-01-01

Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

International Nuclear Information System (INIS)

Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

2000-04-01

As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed

Pending issues for severe accident management in Wolsong plants

International Nuclear Information System (INIS)

Song, Y.M.; Kim, D.H.; Park, S.Y.

2015-01-01

While the fraction of electric power supplied from a PHWR is more than 10% in Korea, the establishment of PHWR safety enhancement based on the SAM (Severe Accident Management) technology is still weak. The final approval on the extended operation and a stress test of Wolsong-1 were made under the condition that SAM is to be enhanced. Under this situation, the current research at KAERI of Korea has a vision to strengthen the unique value of a PHWR by resolving the pending SAM issues devaluating the PHWRs’ original value. Research activities in this area will be presented. This presentation will include: The operating strategy of CFVS (Containment Filtered Vent System) for Wolsong in which vent size and closure pressure are treated because some peak spikes (at failure times of calandria and calandria vault) are difficult to be controlled; Reactor Building failure pressure at which failure probability is treated for different modes such as global and leak failures; the adequacy of DCRV (Degasser Condenser tank Relief Valve) steam relief capacity with severe SGTR source term, and Hydrogen generation and control issue which is specific to CANDU. Furthermore, current SAM guidance has a lack of information on accident diagnostic and prognostic analyses, which is difficult for the TSC (Technical Service Center) emergency staff members to deal with under real accident conditions. Thus, prototypic technologies (such as an accident inferring engine and simulator) together with SAM updates are being developed as key elements to SAM supporting tools called SAMEX-CANDU

Quality assurance in developing countries, with particular reference to Indian experience

International Nuclear Information System (INIS)

Balaramamoorthy, K.; Rao, V.S.G.; Kulkarni, P.G.

1982-01-01

The construction of nuclear power plants is one of the most advanced engineering activities. In India, nuclear power stations are designed, constructed, commissioned, operated and owned by the Department of Atomic Energy. Great emphasis has been put on achieving self-sufficiency in the various activities of the programme, including the manufacture of critical components, such as calandria, heat exchangers and steam generators, and structural components like Zircaloy calandria tubes, pressure tubes and fuel bundles. But like all pioneering ventures this development is not without its problems, costs and delays. In developing countries, local industrial participation is not only desirable but sometimes indispensable and essential. Establishing a nuclear programme very often symbolizes the introduction of modern science and technology, both fundamentally and as applied. To be fully effective, quality assurance (QA) must be comprehensive and cover all engineering activities, including design, procurement, manufacture, construction, commissioning and operation. In the Indian nuclear power programme, well-established and proven QA procedures and practices are being followed during the various stages of nuclear power plant construction and operation. Since training of scientific and technical manpower is an important activity in a developing country, a manpower training programme was initiated in India several years before the introduction of nuclear power plants. The paper deals in detail with the practices of establishing and implementing QA over the years, including training of requisite manpower, problems faced and how they were resolved through progressive indigenization in the manufacture of components, and in construction and operation of nuclear power plants

Experimental study of poison moderator interface movement for shut down system #2(SDS#2) of 540 MWe PHWR

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.; Chawan, D.B.; Ananthan, P.; Sharma, B.S.V.G.; Mohan, L.R.

2005-03-01

The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface, termed as poison moderator interface (PMI). During normal operation of the reactor, the interface moves towards the calandria, mainly because of molecular diffusion from poison to moderator. Other reasons for movement are mixing of poison and moderator due to physical disturbances in the moderator level and to some extent due to temperature difference between the two liquids. The electrical conductivity of these liquids was found to be the most reliable parameter indicating interface movement. For this purpose, two on-line high-pressure conductivity probes have been installed on moderator side for each one of the six poison tanks. During normal operation of reactor, the interface moves slowly towards the calandria over a period of time and gives rise to increase in conductivity. To study the interface pattern and factors affecting the same, a full-scale experimental setup was developed and series of experiments carried out. The experimental results showed that the interface is quite stable and annunciation can be placed around 100 micro siemens/cm before back flushing is initiated. One dimensional diffusion analysis of the obtained experimental data showed that the derived model for PMI setup with diffusion parameter of 900 cm 2 /hr is able to predict the interface movement quite satisfactorily. This report gives an insight into the experiments carried out for estimation of the effective diffusion parameter for the poison moderator interface, model formulation and its prognostic behavior. (author)

Remotely controlled repairs at Douglas Point NGS

International Nuclear Information System (INIS)

Broad, Les

In September, 1977, leakage of heavy water at a rate of 125 kg/hr was detected in an area of the Douglas Point NGS reactor vault below the calandria known as the lower labyrinth. Radiation in the area ranges up to 5000 R/hr and the only ready access was through four 75 mm inspection ports that open into the moderator room. Remote-controlled equipment was designed and built to diagnose the problems and carry out repairs. All damaged piping was fixed, supports were replaced as needed, and system vibration was reduced. The work was done with no injuries and little radiation dose

Consequences assessment for fuel channel failure with consequential moderator drain

International Nuclear Information System (INIS)

Wahba, N.N.; Bayoumi, M.H.

2002-01-01

This paper documents the consequences of spontaneous pressure tube/consequential calandria tube rupture followed by the ejection of end fittings (as a result of guillotine failure of pressure tube) leading to the drain of the moderator. The event is postulated to occur in conjunction with an independent failure of Emergency Coolant Injection System (ECIS). The results of the detailed consequence assessments are used to propose a course of action to mitigate the consequences of such an event. A methodology based on a lumped-parameter model was developed to assess the consequences of the postulated event. (author)

Atomic power project, Kakrapar, Gujarat

International Nuclear Information System (INIS)

Varadarajan, G.

1992-01-01

The atomic power project at Kakrapar, comprising of two units of 235 MW each, went critical very recently in September 1992. The work consisted of construction of reactor and turbine buildings, outer and inner containment walls, calandria vault, natural draught cooling tower, etc. Nearly 152,000m 3 of normal aggregate concrete and 3,500m 3 of heavy aggregate concrete were produced and poured. The paper describes salient innovative construction features of the project. Incidentally, the project received a Certificate of Merit in the Excellence in Concrete competition held by the Maharashtra India Chapter of the American Concrete Institute. (author). 7 figs

Advantages of butterfly valves for power plants

International Nuclear Information System (INIS)

Lapadat, J.T.

1977-01-01

Butterfly valves are increasingly used in nuclear power plants. They are used in CANDU reactors for class 2 and 3 service, to provide emergency and tight shutoff valves for all inlets and outlets of heat exchangers and all calandria penetrations. Guidelines for meeting nuclear power plant valve specifications are set out in ASME Section 3, Nuclear Power Plant Components. Some details of materials of construction, type of actuator, etc., for various classes of nuclear service are tabulated in the present article. The 'fishtail' butterfly valve is an improved design with reduced drag, as is illustrated and explained. (N.D.H.)

Industry partnership: adding value to nuclear refurbishment and maintenance

International Nuclear Information System (INIS)

Gibbins, T.; Bains, N.; Morikawa, D.

2008-01-01

The Point Lepreau Generating Station was the first CANDU 6 unit to be licensed for operation, beginning commercial operation in 1983. It is now become the first CANDU 6 to undergo full refurbishment. As part of the overall project, all 380 fuel channels and associated feeders will be removed and replaced. In order to undertake this project, it was necessary for AECL to design and develop over fifty 'first-of-a-kind tools' for fuel channel and calandria tube replacement. This paper outlines the complexity of the retube tooling project and the industry partnership strategy for the engineered tooling systems development. (author)

Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

International Nuclear Information System (INIS)

Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

2013-01-01

Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

Improvement of top shield analysis technology for CANDU 6 reactor

International Nuclear Information System (INIS)

Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

1996-07-01

As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

Life extension of CANDU reactor cores

International Nuclear Information System (INIS)

Millard, J.; Kerker, J.; Albert, M.

2011-01-01

Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

An experimental study of a flashing-driven CANDU moderator cooling system

Energy Technology Data Exchange (ETDEWEB)

Khartabil, H F; Spinks, N J [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

The results of an experimental study to investigate the feasibility of using a passive flashing-driven natural circulation loop for CANDU-reactor moderator heat rejection are presented. A scaled loop was constructed and tested at conditions approximating those of a CANDU calandria cooling system. The results showed that stable loop operation was possible at simulated powers approaching normal full power. At lower powers, flow oscillations occurred as the flow in the hot-leg periodically changed from two-phase to single-phase. The results from earlier numerical predictions using the CATHENA thermalhydraulics code showed good qualitative agreement with the experimental results. (author). 6 refs., 11 figs.

Eddy current and ultrasonic fuel channel inspection at Karachi Nuclear Power Plant

International Nuclear Information System (INIS)

Mayo, W.R.; Alam, M.M.

1997-01-01

In November of 1993 and in-service inspection was performed on eight fuel channels in the Karachi Nuclear Power Plant (KANUPP) reactor. The workscope included ultrasonic and eddy current volumetric examinations, and eddy current measurement of pressure-to calandria tube gap. This paper briefly discusses the planning strategy of the ultrasonic and eddy current examinations, and describes the equipment developed to meet the requirements, followed by details of the actual channel inspection campaign. The presented nondestructive examinations assisted in determining fitness for service of KANUPP reactor channels in general, and confirmed that the problems associated with channel G12 were not generic in nature. (author)

Assessment of leak detection capability of Candu 6 annulus gas system using moisture injection tests

International Nuclear Information System (INIS)

Nho, Ki Man; Kim, Wang Bae

1998-01-01

The Candu 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside calandria tube and the annulus between these tubes, which forms a closed loop with CO 2 gas recirculating, is called the Annulus Gas System (AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tbe rupture incident. To judge whether the operator action time is enough or not in the design of Wolsung 2, 3, and 4, the Leak Before Break (LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsung Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for dew point rate of rise of Wolsung Unit 2. It was found that the response of the dew point depends on the moisture injection rate, CO 2 gas flow rate and the leak location. The test showed that Candu 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS CO 2 flow rate is approximated. (author)

High-temperature transient creep properties of CANDU pressure tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Chow, C.K.

2002-06-01

During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

The status of improved pressurized heavy water reactor development - A new option for PHWR -

International Nuclear Information System (INIS)

Park, Tae Keun; Yeo, Ji Won

1996-03-01

Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

Contribution to the radioactivity of qatrani Area, western desert, Egypt

International Nuclear Information System (INIS)

El-Shazly, E.M.; El-Sokkary, A.A.; Hussein, H.A.

1974-01-01

The report presents the results of the preliminary theoretical study of a 50 MW(th) heavy water-natural uranium reactor which serves as a heat source for a desalination plant. The calculations were performed by a simple method suggested by the authors that is sufficiently accurate for design purposes. In the report the following relationships are given: dependence of the effective multiplication constant on the lattice pitch and on thickness of the air gap between the pressure and calandria tubes; the lifetime at full-scale operation as function of the core size; the temperature dependence of the effective multiplication constant. The results of control system estimations and power distribution for the core with different reflectors are also provided

Moderator heat recovery of CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Ahmed, S.T.

1986-01-01

A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

Research reactor put Canada in the nuclear big time

International Nuclear Information System (INIS)

Anon.

1988-01-01

The history of the NRX reactor is briefly recounted. When NRX started up in 1947, it was the most powerful neutron source in the world. It is now the oldest research reactor still operating. NRX had to be rebuilt after an accident in 1952, and its calandria was changed again in 1970. Loops in NRX were used to test fuel for the Nautilus submarine, and the first zircaloy pressure tube in the world. At the present time, NRX is in a 'hot standby' condition as a backup to the NRU reactor, which is used mainly for isotope production. NRX will be decommissioned after completion and startup of the new MAPLE-X reactor

First Research Coordination Meeting on Prediction of Axial and Radial Creep in HWR Pressure Tubes. Presentations

International Nuclear Information System (INIS)

2013-01-01

Pressure tube deformation is a critical aging issue in operating Heavy Water Reactors (HWRs). According to the service year, horizontal pressure tubes have three kinds of deformation: diametral creep leading to the flow bypass and the penalty to critical heat flux for fuel rods, longitudinal creep leading to the interference of feeder pipes and/or with fuelling machine, and sagging leading to the interference with in-core components and potential contact between the pressure tube and calandria tube. The CRP scope includes the establishment of a database for pressure tube deformation, microstructure characterization of pressure tube materials collected from HWRs currently operating in Member States and development of a prediction model for pressure tube deformation

Control system design for a 100 MW(th) research reactor

International Nuclear Information System (INIS)

Seshadri, S.N.; Ranganath, M.V.; Singh, Manjit.

1983-01-01

This paper presents the computer simulation carried out to evolve a suitable analog controller for a 100 MW(th) heavy water moderated research reactor under construction at Trombay. The control action is based on the average neutron flux in the reactor core and the reactivity is controlled by adjusting the moderator level in the calandria. A dual control scheme controlling the inflow as well as the outflow was adopted in order to fully exploit the capabilities of control elements. For reasons of reliability, the system consists of three identical channels enabling safe operation even under one channel failure. Based on the simulation studies a suitable compensation network was incorporated to achieve satisfactory system response. (author)

Examination of core components removed from CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

1988-11-01

Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

Design features of ACR in severe accident mitigation

International Nuclear Information System (INIS)

Shapiro, H.; Krishnan, V.S.; Santamaura, P.; Lekakh, B.; Blahnik, C.

2007-01-01

New reactor designs require the evaluation of design alternatives to reduce the radiological risk by preventing severe accidents or by limiting releases from the plant in the event of such accidents. The Advanced CANDU Reactor TM (ACR TM ) design has provisions to prevent and mitigate severe accidents. This paper describes key ACR design features for severe accident mitigation. It provides a high-level overview of the findings to date. Several design provisions have not yet been finalized or decided, but the designers are keenly aware of the SAM concepts and their requirements. The active heat sinks for 'vessels' (i.e., the fuel channels, the calandria vessel, the calandria end-shields and the calandria vault) are all amply capable of dissipating the severe accident heat loads. These heat sinks are designed to be operable under severe accident environmental conditions; however, their operability is yet to be confirmed by assessments. The active heat sinks for the various process vessels are 'backed up' by passive heat sinks (i.e., steaming plus water make-up from the RWS). The supply side of passive heat sinks is simple, rugged, and not vulnerable to failures of plant systems. The importance of the steam relief side is recognized, and the adequate relief capacity will be provided. The passive heat sinks will give the SAM more than 1 day (likely several days) to diagnose the accident and to establish the ultimate heat sinks. The spray system for containment pressure suppression is designed for high reliability and has ample capacity to ensure low containment leakage without external intervention, after which time alternative supply to the sprays can be brought on line manually. The sprays are backed up by the LACs which are assessed for operability following a severe accident. The strong ACR containment will provide a long time of completely passive protection for any severe accident at decay power. Its characteristics are not prone to catastrophic failures. The

A three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

International Nuclear Information System (INIS)

Bang, K. H.; Lee, J. Y.; Yoo, S. O.; Kim, M. W.; Kim, H. J.

2002-01-01

Three-dimensional analyses of fluid flow and heat transfer has been performed in this study. The simulation of SPEL experimental work and comparison with experimental data has been carried out to verify the analyses models. Moreover, to verify the CANDU-6 reactor type, analyses of fluid flow and heat transfer in the calandria under the condition of steady state has been performed using FLUENT code, which is the conventional code for a three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR thermal-hydraulics. It is found that the maximum temperature in the moderator is 347K (74 ), so that the moderator has the enough subcoolability to ensure the integrity of pressure tube during LOCA conditions

Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

International Nuclear Information System (INIS)

Fath, H.E.S.

1988-01-01

Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

Asset management program

International Nuclear Information System (INIS)

Wison, P.; Newman, G.

2013-01-01

In order to understand our assets we have been assessing the condition of the units in our nuclear power plants developing asset life management options on a component by component basis. We have concluded that with the right work and planning we will be able to manage the units in a way that balances capacity requirements over the long term and at the same time manage the demand on critical resources. Major component replacement outages include Installing/removing bulkheads, pressure tube and calandria tube replacement, feeder replacement, steam generator replacement, supporting facilities and infrastructure, reactor inspections and maintenance including tooling enhancements, additional non reactor systems inspection & testing and continued research and analysis. These plans will have to take into account cost, resource and capacity requirements.

The generation of calandria tube (CT) inner diameter profiles from fuel channel (FC) inspection data

Energy Technology Data Exchange (ETDEWEB)

Sedran, P.J., E-mail: paul.sedran@amec.com [AMEC NSS, Toronto, ON (Canada); Rankin, B., E-mail: brankin@nbpower.com [NB Power, Fredericton, NB (Canada); Lemire, C., E-mail: Lemire.Christian@hydro.qc.ca [Hydro-Quebec, Montreal, QC (Canada)

2015-07-01

Studies of CT deformation at spacer locations, key to the development of FC deformation modelling, have been limited by the availability of gauging measurements from removed CTs. In [1], it was proposed that CT dimensional profiles could be generated using FC inspection data. Since then, the concept was investigated further by assessing: (1) the normalisation of gap measurements to the diameter of the spacer coil, (2) the validity of gap measurements from inspections of Point Lepreau and Gentilly-2, and the CT dimensional profiles generated from the inspection data.It was concluded, from the work presented in this paper, that the CT-PT gap data and the CT dimensional profiles generated using the data from the two subject inspections are reasonable. (author)

Fuel channel design improvements for large CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

Absorber materials in CANDU PHWR's

International Nuclear Information System (INIS)

Price, E.G.; Boss, C.R.; Novak, W.Z.; Fong, R.W.L.

1995-03-01

In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in a relatively benign environment of low pressure, low temperature heavy water between neighbouring rows of columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a designed back-fit resolved the problem. (author). 3 refs., 1

Severe core damage experiments and analysis for CANDU applications

International Nuclear Information System (INIS)

Mathew, P.M.; White, A.J.; Snell, V.G.; Bonechi, M.

2003-01-01

AECL uses the MAAP CANDU code to calculate the progression of a severe core damage accident in a CANDU reactor to support Level 2 Probabilistic Safety Assessment and Severe Accident Management activities. Experimental data are required to ensure that the core damage models used in MAAP CANDU code are adequate. In SMiRT 16, details of single channel experiments were presented to elucidate the mechanisms of core debris formation. This paper presents the progress made in severe core damage experiments since then using single channels in an inert atmosphere and results of the model development work to support the experiments. The core disassembly experiments are conducted with one-fifth scale channels made of Zr-2.5wt%Nb containing twelve simulated fuel bundles in an inert atmosphere. The reference fuel channel geometry consists of a pressure tube/calandria tube composite, with the pressure tube ballooned into circumferential contact with the calandria tube. Experimental results from single channel tests showed the development of time-dependent sag when the reference channel temperature exceeded 850 degC. The test results also showed significant strain localization in the gap at the bundle junctions along the bottom side of the channel, thus suggesting creep to be the main deformation mechanism for debris formation. An ABAQUS finite element model using two-dimensional beam elements with circular cross-section was developed to explain the experimental findings. A comparison of the calculated central sag (at mid-span), the axial displacement at the free end of the channel and the post-test sag profile showed good agreement with the experiments, when strain localization was included in the model, suggesting such a simple modelling approach would be adequate to explain the test findings. The results of the tests are important not only in the context of the validation of the analytical tools and models adopted by AECL for the severe accident analysis of CANDU reactors but

Simulating the behaviour of zirconium-alloy components in nuclear reactors

International Nuclear Information System (INIS)

Coleman, C.E.

2001-12-01

To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

Application of gas shielded arc welding and submerged arc welding for fabrication of nuclear reactor vessels

International Nuclear Information System (INIS)

Gehani, M.L.; Rodrigues, W.D.

1976-01-01

The remarkable progress made in the development of knowhow and expertise in the manufacture of equipment for nuclear power plants in India is outlined. Some of the specific advances made in the application of higher efficiency weld processes for fabrication of nuclear reactor vessels and the higher level of quality attained are discussed in detail. Modifications and developments in submerged arc, gas tungsten arc and gas metal arc processes for welding of Calandria which have been a highly challenging and rewarding experience are discussed. Future scope for making the gas metal arc process more economical by using various gas-mixes like Agron + Oxygen, Argon + Carbon Dioxide, Argon + Nitrogen (for Copper Alloys) etc., in various proportions are outlined. Quality and dimensional control exercised in these jobs of high precision are highlighted. (K.B.)

Remote field eddy current testing

International Nuclear Information System (INIS)

Cheong, Y. M.; Jung, H. K.; Huh, H.; Lee, Y. S.; Shim, C. M.

2001-03-01

The state-of-art technology of the remote field eddy current, which is actively developed as an electromagnetic non-destructive testing tool for ferromagnetic tubes, is described. The historical background and recent R and D activities of remote-field eddy current technology are explained including the theoretical development of remote field eddy current, such as analytical and numerical approach, and the results of finite element analysis. The influencing factors for actual applications, such as the effect of frequency, magnetic permeability, receiving sensitivity, and difficulties of detection and classification of defects are also described. Finally, two examples of actual application, 1) the gap measurement between pressure tubes and calandria tube in CANDU reactor and, 2) the detection of defects in the ferromagnetic heat exchanger tubes, are described. The future research efforts are also included

Assessment of moderator integrity using realisitc model and parametric studies on thermal-hydraulic characteristics in SPEL

International Nuclear Information System (INIS)

Yoo, S. H.; Kim, M. W.; Kang, S. C.; Kim, H. J.; Min, B. J.; Yoon, C.

2002-01-01

Three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL(Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. In this study, three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. Moreover, the behavior of temperature distribution inside calandria has also been investigated, when the flowrate of moderator is changed with time

Radiolytic generation of gases in reactors

International Nuclear Information System (INIS)

Ramshesh, V.; Venkateswarlu, K.S.

1988-01-01

Water or heavy water is used in different circuits in a reactor. Their most common use is as a moderator and/or as a coolant. Light water is used at other places such as in end shield, calandria vault etc., In the process they are exposed to intense ionizing radiation and undergo radiolytic degradation. The molecular produts of radiolysis are hydrogen, hydrogen peroxide and oxygen. As is commonly known if hydrogen is formed beyond a certain level, in the presence of oxygen it may lead to combustion or even explosion. Thus one should comprehend the basic principles of radiolysis and see whether the concentration of these gases under various conditions can be worked out. This report attempts to analyse in depth the radiolytic generation of gases in reactor systems. (author). 3 tabs

Performance evaluation of eddy current transducers and associated instrumentation of integrated garter spring repositioning system

International Nuclear Information System (INIS)

Sharma, B.S.V.G.; Shyam, T.V.; Shrivastava, A.K.; Sinha, R.K.

1997-01-01

To extend the life of coolant channels of operating Indian Pressurised Heavy Water Reactors (PHWRs) of an early generation, repositioning of dislocated Garter Spring (GS) spacers is necessary. For this purpose a remotely operated system named INtegrated Garter spring REpositioning System (INGRES) has been developed. As a part of this system, eddy current transducers namely Garter Spring Detection Probe (GSDP) and Concentricity Detection Probe (CDP) along with respective signal processor units have been designed and developed. These devices detect GS spacers and eccentricity between Pressure Tube (PT) and Calandria Tube (CT) of the channel respectively. During a recent campaign of INGRES at Madras Atomic Power Station unit-2 (MAPS-2), these transducer systems have fulfilled intended design and operational objectives besides providing additional information regarding channel. These aspects are discussed. (author). 6 figs

Design of improved detection instrumentation for the annulus gas system for wolsong 2

International Nuclear Information System (INIS)

Kim, Seog Nam; Koo, Jun Mo; Chang, Ik Ho; Jung, Ho Chang; Han, Sang Joon

1996-01-01

The improved and advanced Annulus Gas System (AGS) has been developed for Wolsong 2 to satisfy the requirements of the regulatory body. The Atomic Energy Control Board (AECB) required a shorter detection time following a small leak from a pressure tube and/or calandria tube. This paper describes licensing requirements, functional requirements and detail design description for the AGS. The Wolsong unit No. 1 AGS was designed to operate as a stagnant system normally requiring only pressure regulation and having provisions for purging. The improved AGS involves the adoption of gas recirculation in AGS, duplication of dew point indicators with additional instrumentation and sampling provisions to prompt operator action. The improved system operates in the recirculation mode with continuous dew point measurement for leak detection. An AGS with improved detection instrumentation is provided. 8 refs., 3 figs. (author)

Technology development for special nuclear components

International Nuclear Information System (INIS)

Sanatkumar, A.

1994-01-01

One of the attractive features of Candu Pressurised Heavy Water Reactor design which influenced the decision to make it the foundation of our nuclear power programme, is that its main components (calandria, end shields, coolant channel components) are relatively simple - in comparison with reactor pressure vessel and associated components of Boiling Water Reactors or Pressurised Water Reactors - and considered to be within the scope of manufacture of developing countries. Over the last two decades, India has been very successful in technology development in many important and critical areas. We are now about to launch the construction of the first 500 MWe PHWR project at Tarapur. In this context, this paper focuses attention on some of the aspects relating to self-reliance in design, engineering and manufacture of these special components as currently perceived. (author). 3 refs

Moderator clean-up system in a heavy water reactor

International Nuclear Information System (INIS)

Sasada, Yasuhiro; Hamamura, Kenji.

1983-01-01

Purpose: To decrease the fluctuation of the poison concentration in heavy water moderator due to a heavy water clean-up system. Constitution: To a calandria tank filled with heavy water as poison-containing moderators, are connected both end of a pipeway through which heavy water flows and to which a clean-up device is provided. Strongly basic resin is filled within the clean-up device and a cooler is disposed to a pipeway at the upstream of the clean-up device. In this structure, the temperature of heavy water at the inlet of the clean-up device at a constant level between the temperature at the exit of the cooler and the lowest temperature for the moderator to thereby decrease the fluctuation in the poison concentration in the heavy water moderator due to the heavy water clean-up device. (Moriyama, K.)

MAPLE: a Canadian multipurpose reactor concept for national nuclear development

International Nuclear Information System (INIS)

Lidstone, R.F.

1984-06-01

Atomic Energy of Canada Limited, following an investigation of Canadian and international needs and world-market prospects for research reactors, has developed a new multipurpose concept, called MAPLE (Multipurpose Applied Physics Lattice Experimental). The MAPLE concept combines H 2 O- and D 2 O-moderated lattices within a D 2 O calandria tank in order to achieve the flux advantages of a basic H 2 O-cooled and moderated core along with the flexibility and space of a D 2 O-moderated core. The SUGAR (Slowpoke Uprated for General Applied Research) MAPLE version of the conept provides a range of utilization that is well suited to the needs of countries with nuclear programs at an early stage. The higher power MAPLE version furnishes high neutron flux levels and the variety of irradiation facilities that are appropriate for more advanced nuclear programs

Tight fitting garter springs-MODAR

Energy Technology Data Exchange (ETDEWEB)

Kazimer, D. [Bruce Power, Tiverton, Ontario (Canada)

2011-07-01

Annulus spacers are used in CANDU reactors to maintain the annular gap between two tubes - an inner pressure tube (PT) and the outer calandria tube (CT). Typically four annulus spacers are used in one fuel channel assembly, each at a specified axial position. Bruce Unit 8 and many other CANDU units were constructed with tight-fitting garter springs (TFGS). The TFGS were not designed to be detected or relocated by the conventional tool, Spacer Location And Repositioning (SLAR) processes. Due to non-optimal 'As Left' construction locations for the Bruce Unit 8 TFGS, PT/CT contact has been predicted to occur well prior to its End of Life (EOL). Bruce Power entered a Project with AECL-CRL to design, manufacture and test and implement a new tooling system that would detect and reposition tight fitting annulus spacers. (author)

Radiological experience on decontamination of moderator and associated system at NAPS-1

International Nuclear Information System (INIS)

Yadav, C.L.R.; Mitra, S.R.; Pawar, S.K.; Lal Chand

2000-01-01

Narora Atomic Power Station, the first of Indian standardized Pressurized Heavy Water Reactor, is faced with a problem of 60 Co contamination in moderator and its associated system. This contamination has resulted in increase of collective dose contribution. As a part of ALARA campaign in NAPS it was decided to decontaminate the moderator and associated systems and also incorporate modification which will further control the 60 Co contamination. As a part of decontamination program several experiments to determine the effectiveness of the chemical formulation on SS and cupro-nickel surfaces were carried out on various moderator system equipment before finalizing the formulation for full scale decontamination of moderator system. This paper gives an overview of various modifications in system and decontamination efficiency of various chemical formulation which were used for decontamination of moderator system (excluding calandria) and associated equipment. (author)

Inspection of Candu Nuclear Reactor Fuel Channels

International Nuclear Information System (INIS)

Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

1986-01-01

The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author) [pt

Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

International Nuclear Information System (INIS)

Kalidas, R.

2005-01-01

Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

2015-05-15

The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

International Nuclear Information System (INIS)

Im, Sunghyuk; Sung, Hyung Jin; Seo, Han; Bang, In Cheol; Kim, Hyoung Tae

2015-01-01

The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Evaluation of the mechanical properties of SA 333 Gr.6, AISI 304 and Zr-2.5% Nb through Automated Ball Indentation (ABI) technique

International Nuclear Information System (INIS)

Balakrishnan, K.S.; Rath, B.N.; Shriwastaw, R.S.; Ramadasan, E.; Kulkarni, R.V.; Sahoo, K.C.

2009-08-01

Automated Ball Indentation (ABI) technique has been employed in evaluating the tensile property data on three materials, namely SA333 Gr.6 carbon steel (used as PHT piping), AISI 304 (used as calandria vessel) and Zr-2.5% Nb (used as coolant tube) in Pressurised Heavy Water Reactors (PHWRs) with a view to exploring the applicability of ABI technique in providing reliable mechanical property data. The exercise was carried out in cooperation with a second laboratory where conventional tension tests alone were conducted such that the output of the study could be independently monitored and evaluated in an unbiased manner. The results generated in the authors' laboratory were found to be fully in agreement with what were obtained through conventional tension tests. Thus the study has been successful in establishing the reliability of the data obtained through miniature route especially in the case of coolant tube which has immense applications. (author)

Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

International Nuclear Information System (INIS)

Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

2008-01-01

Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

International Nuclear Information System (INIS)

Nawathe, Shirish; Umashankari, P.; Balakrishnan, Kamala; Mahajan, S.C.; Kakodkar, A.

1991-01-01

A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

A computer model for hydride blister growth in zirconium alloys

International Nuclear Information System (INIS)

White, A.J.; Sawatzky, A.; Woo, C.H.

1985-06-01

The failure of a Zircaloy-2 pressure tube in the Pickering unit 2 reactor started at a series of zirconium hydride blisters on the outside of the pressure tube. These blisters resulted from the thermal diffusion of hydrogen to the cooler regions of the pressure tube. In this report the physics of thermal diffusion of hydrogen in zirconium is reviewed and a computer model for blister growth in two-dimensional Cartesian geometry is described. The model is used to show that the blister-growth rate in a two-phase zirconium/zirconium-hydride region does not depend on the initial hydrogen concentration nor on the hydrogen pick-up rate, and that for a fixed far-field temperature there is an optimum pressure-type/calandria-tube contact temperature for growing blisters. The model described here can also be used to study large-scale effects, such as hydrogen-depletion zones around hydride blisters

A fluid-solid finite element method for the analysis of reactor safety problems

International Nuclear Information System (INIS)

Mitra, Santanu; Kumar, Ashutosh; Sinhamahapatra, K.P.

2006-01-01

The work presented herein can broadly be categorized as a fluid-structure interaction problem. The response of a circular cylindrical structure subjected to cross flow is examined using the finite element method for both the liquid and the structure domains. The cylindrical tube is mounted elastically at the ends and is free to move under the action of the unsteady flow-induced forces. The fluid is considered to be acoustic compressible and viscous. A Galerkin finite element method implemented on a triangular mesh is used to solve the time-dependent Navier-Stokes equations. The cylinder motion is modeled using a five-degrees of freedom generalized shell element structural dynamics model. The numerical simulations of the response of the calandria tubes/pressure tubes, adjustor rod and shut-off rod of a nuclear reactor are presented. A few typical results are presented to assess the accuracy and applicability of the developed modules

Dynamic analysis of a reactor building on alluvial soil

International Nuclear Information System (INIS)

Arya, A.S.; Chandrasekaran, A.R.; Paul, D.K.; Warudkar, A.S.

1977-01-01

The reactor building consists of reinforced concrete internal framed structure enclosed in double containment shells of prestressed and reinforced concrete all resting on a common massive raft. The external cylindrical shell is capped by a spherical dome while the internal shell carries a cellular gird slab. The building is partially buried under ground. The soil consists of alluvial going to 1000 m depth. The site lies in a moderate seismic zone. The paper presents the dynamic analysis of the building including soil-structure interaction. The mathematical model consists of four parallel, suitably interconnected struxtures, namely inner containment, outer containment, internal frame and the calandria vault. Each one of the parallel structures consists of lumped-mass beam elements. The soil below the raft and on the sides of outer containment shell is represented by elastic springs in both horizontal and vertical directions. The various assumpions required to be made in developing the mathematical model are briefly discussed in the paper. (Auth.)

Monte Carlo-based validation of the ENDF/MC2-II/SDX cell homogenization path

International Nuclear Information System (INIS)

Wade, D.C.

1979-04-01

The results are presented of a program of validation of the unit cell homogenization prescriptions and codes used for the analysis of Zero Power Reactor (ZPR) fast breeder reactor critical experiments. The ZPR drawer loading patterns comprise both plate type and pin-calandria type unit cells. A prescription is used to convert the three dimensional physical geometry of the drawer loadings into one dimensional calculational models. The ETOE-II/MC 2 -II/SDX code sequence is used to transform ENDF/B basic nuclear data into unit cell average broad group cross sections based on the 1D models. Cell average, broad group anisotropic diffusion coefficients are generated using the methods of Benoist or of Gelbard. The resulting broad (approx. 10 to 30) group parameters are used in multigroup diffusion and S/sub n/ transport calculations of full core XY or RZ models which employ smeared atom densities to represent the contents of the unit cells

Monte Carlo; based validation of the ENDF/MC2-II/SDX cell homogenization path

International Nuclear Information System (INIS)

Wade, D.C.

1978-11-01

The results are summarized of a program of validation of the unit cell homogenization prescriptions and codes used for the analysis of Zero Power Reactor (ZPR) fast breeder reactor critical experiments. The ZPR drawer loading patterns comprise both plate type and pin-calandria type unit cells. A prescription is used to convert the three dimensional physical geometry of the drawer loadings into one dimensional calculational models. The ETOE-II/MC 2 -II/SDX code sequence is used to transform ENDF/B basic nuclear data into unit cell average broad group cross sections based on the 1D models. Cell average, broad group anisotropic diffusion coefficients are generated using the methods of Benoist or of Gelbard. The resulting broad (approx. 10 to 30) group parameters are used in multigroup diffusion and S/sub n/ transport calculations of full core XY or RZ models which employ smeared atom densities to represent the contents of the unit cells

Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

International Nuclear Information System (INIS)

Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

2001-01-01

Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

Mode shape and natural frequency identification for seismic analysis from background vibration

International Nuclear Information System (INIS)

Bhan, S.; Wozniak, Z.

1986-10-01

Background vibration in a CANDU plant can be used to determine the dynamic characteristics of major items of equipment, such as calandria, the fuelling machines and the primary heat transport pumps. These dynamic characteristics can then be used to verify the seismic response of the equipment which, at present, is based on theoretical models only. The feasibility and basic theory of this new approach (which uses accelerations measured at several points on a structure and does not require knowledge of the source of excitation) was established in Phase I of the study. This report is based on Phase II in which the methods of analysis developed in Phase I were improved and verified experimentally. A Fast Fourier Transform (FFT) algorithm was incorporated and an interactive curve fitting technique was developed to obtain the dynamic characteristics in the form of natural frequencies, mode shapes and damping ratios. The method is now available for use at a CANDU plant

Heat transfer study of a submerged reactor channel under boil-off condition

Energy Technology Data Exchange (ETDEWEB)

Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

2012-12-15

Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

Size determinations, by ultrasonic techniques, of cracks in hydride blisters formed in Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Trujillo Badillo, Giovanna; Desimone, Carlos; Domizzi, Gladys

1999-01-01

Non destructive techniques (NDT) are very useful in the detection of flaws produced in structural components in service. During the service of CANDU nuclear power reactors, it is possible that pressure tubes (PT) may contact calandria tubes (CT). After the PT/CT contact, zirconium hydride blisters may form at the point of contact depending on the concentration of hydrogen/deuterium. Zirconium hydride is brittle and is therefore prone to cracking under stress. Ultrasonic NDT is routinely use during PT in service inspection. In order to be able of detecting cracked blisters, it is of great importance the development of standards to calibrate the employed equipment. On this purpose, hydride blisters were grown, in laboratory, on sections of pressure tube. The cracks in the blisters were detected and measured by ultrasonic techniques. The obtained results were compared with measurements carried out in optic microscope, on successive sections of the samples. The crack tip diffraction technique was found to be the more effective for the mentioned ends. (author)

Assessment of the applicability of AWJ technique for dismantling the reactor of Fugen. Performance of underwater-cutting thick plate and testing of sound-based monitoring for underwater-cutting process

International Nuclear Information System (INIS)

Maruyama, Shin-ichiro; Nishio, Shin-ichi

2010-01-01

The reactor of Fugen is characterized by its double-walled pressure tube construction that is composed of pressure tubes and calandria tubes. The reactor dismantlement has been planning on dismantling it under water and the abrasive water jet (AWJ) underwater-cutting method is chosen as an option among simultaneous double tubes cutting technologies. For assessing the applicability of the AWJ cutting technology, a thick plate was cut under water by the small AWJ cutting machine. In addition, since cutting causes muddiness in water, cutting was monitored by the sound-based monitoring system which was adopted as a secondary cutting monitoring method. As a results, it was demonstrated that one-phase cutting was possible under water for a stainless-steel plate with 150mm thickness and that the relationship between cutting depth and capable cutting speed could be predictable. As for the sound-based cutting monitoring, the predictability whether or not cutting would be successful was verified by checking the change of sounds level. (author)

Separating rings detection in fuel channels of Embalse NPP

International Nuclear Information System (INIS)

Obrutsky, L.S.; Otero, P.A.; Schmidt, O.A.

1988-01-01

The design specifications of Embalse Nuclear Power Plants (CANDU Type Reactor 600Mw) define the positions to be taken by 4 separating rings of the fuel channels. Experience has demonstrated the displacement possibility of the above mentioned rings. It means a risk of contact between pressure tube and calandria tube. In order to determine the position of separating rings, an inspection system based on Eddy Currents technique was developed by CNEA personnel. Detection is performed through two special probes operating according the ''emitter-receiver'' principle. Obtained signals and its relative position are recorded in a video tape and registered in paper. The probe is telecommanded by an automatic equipment. In this paper the construction and calibration of the detection equipment is described, as well as the propulsion. Final results are also outlined in the inspection carried out in November 1986 when an effective displacement of separating rings was verified from its design position in most of the inspected tubes

Molten fuel-moderator interaction

International Nuclear Information System (INIS)

Lee, J.H.S.; Kynstautas, R.

1987-02-01

A critical review of the current understanding of vapor explosions was carried out. It was concluded that, on the basis of actual industrial accidents and large scale experiments, energetic high yield steam explosion cannot be regarded as an improbable event if large quantities of molten fuel and coolant are mixed together. This study also reviewed a hydrodynamic transient model proposed by Henry and Fauske Associates to assess a molten fuel-moderator interaction event. It was found that the proposed model negates a priori the possibility of a violent event, by introducing two assumptions: 1) fine fragmentation of the molten fuel, and ii) rapid heat transfer from the fine fragments to form steam. Using the Hicks and Menzies thermodynamic model, maximum work potential and pressure rise in the calandria were estimated. However, it is recommended that a more representative upper bound model based on an underwater explosion of a pressurized volume of steam be developed

Fabrication of high performance components for Indian nuclear reactors

International Nuclear Information System (INIS)

Jayaraj, R.N.

2011-01-01

Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

CANDU heat sinks improvements as a follow up to Fukushima Daiichi accident ''the regulator perspective''

Energy Technology Data Exchange (ETDEWEB)

Mesmous, Noreddine; Harwood, Chris [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

2015-06-15

The purpose of this paper is to provide a summary of the Canadian Nuclear Safety Commission (CNSC) recommendations related to improving the heat sink strategy as a follow up to the Fukushima Daiichi Accident (FDA). As a follow up to FDA, CNSC staff tasked the Nuclear Power Plant (NPP) licensees to review the lessons learned from the FDA and re-examine the NPP safety cases. The reviews have examined the CANDU defence-in-depth strategy and considered events more severe than those that have historically been regarded as credible, and evaluated their impact on the NPPs safety. Availability of emergency equipment was shown to be crucial during the FDA and its availability could have arrested the accident progression early enough to minimize any radioactive release to the environment. As a result, licensees presented appropriate evaluations of the means to provide coolant make-up to the primary Heat Transport System (HTS), boilers, moderator, calandria vault, and irradiated fuel pools.

Preliminary assessment of an S.G.H.W. type research reactor

International Nuclear Information System (INIS)

Bicevskis, A.; Chapman, A.G.; Hesse, E.W.

1970-08-01

A preliminary design study has been made of a research reactor, based on the enriched S.G.H.W.R. concept, to be used for power reactor fuel irradiation, isotope production, basic research, and training in nuclear technology. A reactor physics assessment established a core size which would allow uninterrupted operation for the required irradiation period consistent with low capital and operating costs. A design was selected with 24 channels, a D 2 O calandria diameter of 2.7 m and an overall core height of 4.0 m. The capital cost was estimated as $750,000 for the fuel and $1,600,000 for the moderator, the refuelling cost being $340,000 per annum. A thermal design study showed that the fission heat of 65 MW could be transmitted to pressurised light water at 200 lb/in 2 abs. and rejected to sea water in two conventional U-tube heat exchangers. The basic design is flexible and can be adapted to meet many special requirements. (author)

Simulation of hydrogen migration and blisters formation in zirconium alloys

International Nuclear Information System (INIS)

Saliba, R.O.

1991-06-01

The phenomenon of hydrogen migration and hydride blister growth after pressure tube/calandria tube contact in CANDU reactors is addressed. This phenomenon is by now regarded as an important factor limiting reactors lifetime, since it originated Pickering incident in 1983. Numerical results of thermally-assisted diffusion in excellent agreement with quasi-analytical solutions of the mathematical model were obtained. A sensitivity analysis was performed to assess the accuracy of these results. Some two-dimensional calculations are also included to demonstrate the capabilities of the numerical methods. The main outcomes of the work are the following: a through understanding of the mathematics and physics involved in hydrogen migration under thermal gradients. The validation of a numerical procedure based on a regularization of the constitutive equations. Blister growth rates in slab geometries for initial concentrations that span the full range of technological interest. Some preliminary two-dimensional results allow the design of future developments. (Author) [es

An improved model to predict nonuniform deformation of Zr-2.5 Nb pressure tubes

International Nuclear Information System (INIS)

Lei, Q.M.; Fan, H.Z.

1997-01-01

Present circular pressure-tube ballooning models in most fuel channel codes assume that the pressure tube remains circular during ballooning. This model provides adequate predictions of pressure-tube ballooning behaviour when the pressure tube (PT) and the calandria tube (CT) are concentric and when a small (<100 degrees C) top-to-bottom circumferential temperature gradient is present on the pressure tube. However, nonconcentric ballooning is expected to occur under certain postulated CANDU (CANada Deuterium Uranium) accident conditions. This circular geometry assumption prevents the model from accurately predicting nonuniform pressure-tube straining and local PT/CT contact when the pressure tube is subjected to a large circumferential temperature gradient and consequently deforms in a noncircular pattern. This paper describes an improved model that predicts noncircular pressure-tube deformation. Use of this model (once fully validated) will reduce uncertainties in the prediction of pressure-tube ballooning during a postulated loss-of-coolant accident (LOCA) in a CANDU reactor. The noncircular deformation model considers a ring or cross-section of a pressure tube with unit axial length to calculate deformation in the radial and circumferential directions. The model keeps track of the thinning of the pressure-tube wall as well as the shape deviation from a reference circle. Such deviation is expressed in a cosine Fourier series for the lateral symmetry case. The coefficients of the series for the first m terms are calculated by solving a set of algebraic equations at each time step. The model also takes into account the effects of pressure-tube sag or bow on ballooning, using an input value of the offset distance between the centre of the calandria tube and the initial centre of the pressure tube for determining the position radius of the pressure tube. One significant improvement realized in using the noncircular deformation model is a more accurate prediction in

Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal-chemical experiment CS28-1, I - Steady state

International Nuclear Information System (INIS)

Rhee, Bo Wook; Kim, Hyoung Tae; Park, Joo Hwan

2008-01-01

To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal-chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H 2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20-25 deg. C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO 2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using 'transparent' assumption for the CO 2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

International Nuclear Information System (INIS)

Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

2008-01-01

Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

The steam generating heavy water reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

Prediction of hydrogen distribution in the reactor building in CANDU6 plant

International Nuclear Information System (INIS)

Jin, Y.; Song, Y.

2008-01-01

The CANDU plants have a lot of zircaloy. The fuel cladding, calandria tubes and pressure tubes are made of zircaloy. The zircaloy can be oxidized and hydrogen is generated during severe accident progression. The detonation or deflagration to detonation transition (DDT) due to hydrogen combustion may occur if the local hydrogen concentration or global hydrogen concentration exceeds certain value. The detonation may result in the rupture of the reactor building. The inside of the reactor building of CANDU plants is complex. So prediction of hydrogen distribution in the reactor building is important. This prediction is made using ISAAC code and GOTHIC code. ISAAC code partitioned the reactor building in to 7 compartments. GOTHIC code modeled the CANDU6 reactor building using 12 nodes. The hydrogen concentrations in the various compartments in the reactor building are compared. GOTHIC code slightly underpredicts hydrogen concentration in the F/M rooms than ISAAC code, but trend is same. The hydrogen concentration in the boiler room and the moderator room shows almost same as for both codes. (author)

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

Development of a flow restrictor for CANDU fuel channels

International Nuclear Information System (INIS)

Schroeter, F.; Antonaccio, C.; Masciotra, H.; Klink, A.

2013-01-01

Due to the creep and neutron growth phenomena experienced by the components inside the reactor during operation of CNE it is expected that both the fuel channels and the Liquid Injection lances increase their permanent deformation. One of the deformation types that these two components experiment is the SAG, which is what happens with any beam supported on their extremes to which a load is applied, except for this case that due to creep and neutronic effect growth, part of the deformation is not elastic and increases with time To solve or avoid this condition, two solutions exist, one is to replace the pressure tubes, forcing the calandria tube to recover to a near to original position or to design a device that permits defueling of the channel without modifying the pressure drop and in this way not to affect the distribution of coolant in the core. In some channels it was decided to replace the pressure tube and in others it was decided to defuel them proposing a design for a flow restrictor. (author

Comparison of analyzed design-basis events to actual plant transients

International Nuclear Information System (INIS)

Geeting, M.W.; Hightower, N.T. III; Fields, C.C.

1992-01-01

Fitness-for-Service Guidelines have recently been developed to provide acceptance criteria and evaluation methods for assessment of the integrity of the Zr-2.5 Nb pressure tubes in operating Canada deuterium uranium (CANDU) reactors. The guidelines provide a methodology for the evaluation of specific conditions in a single tube, such as manufacturing and inservice generated flaws, hydride blisters formed at points of contact between a pressure tube and its calandria tube, and generic degradation of pressure tube properties in service. The guidelines are divided into three sections. The first section describes the requirements that must be met to qualify the tubes ofr continued service. The second section provides the material properties data-base information needed to carry out the assessments. The third section provides the technical basis for the acceptance criteria and evaluation procedures as well as justifications and descriptions of the data bases. The guidelines were issued to CANDU reactor operators for trial use and released to the Atomic Energy Control Board of Canada for review and comment in May 1991

Implementation of Moderator Circulation Test Temperature Measurement System

Energy Technology Data Exchange (ETDEWEB)

Lim, Yeong Muk; Hong, Seok Boong; Kim, Min Seok; Choi, Hwa Rim [KAERI, Daejeon (Korea, Republic of); Kim, Hyung Shin [Chungnam University, Daejeon (Korea, Republic of)

2016-05-15

Moderator Circulation Test(MCT) facility is 1/4 scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. MCT is an equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV(Particle Image Velocimetry) is performed under the iso-thermal test conditions. The Korea Atomic Energy Research Institute (KAERI) started implementation of MCT Temperature Measurement System (TMS) using multiple infrared sensors. To control multiple infrared sensors, MCT TMS is implemented using National Instruments (NI) LabVIEW programming language. The MCT TMS is implemented to measure sensor data of multiple infrared sensors using the LabVIEW. The 35 sensor pipes of MCT TMS are divided into 2 ports to meet the minimum measurement time of 0.2 seconds. The software of MCT TMS is designed using collection function and processing function. The MCT TMS has the function of monitoring the states of multiple infrared sensors. The GUI screen of MCT TMS is composed of sensor pipe categories for user.

Malfunction and failure in regulation and protection systems of PHWRs -identification of deficient areas and suggested scope of improvements

International Nuclear Information System (INIS)

Jain, A.K.; Prabhat Kumar; Arya, R.C.

1997-01-01

The reactor regulation and protection systems have changed significantly in Narora Atomic Power Station type of reactors as compared to RAPS design. As compared to total negative reactivity worth offered by moderator dump for reactor shutdown in Rajasthan Atomic Power Station, the worth of the two fast acting shutdown systems (PSS and SSS) is far lesser. In fact the worth is not even adequate to maintain the shutdown indefinitely. This has necessitated incorporation of two slow acting systems for reactor regulation and protection i.e. ALPS and GRAB systems. The negative reactivity insertion rate from fast acting PSS and SSS is however much faster as compared to moderator dump. The above changes have brought about significant changes in reactivity devices and also associated problems due to larger number of systems and equipment. The layout has also become very congested on the top of calandria vault making maintenance and inspection more cumbersome. The paper highlights the malfunction and deficiencies observed in PSS, SSS and regulating rods and identifies the further scope of improvement. (author)

Dynamic strain analysis of structures employing digital signal processing, storage and display

Energy Technology Data Exchange (ETDEWEB)

Patwardhan, P K; Misra, V M; Kumar, Surendra

1975-01-01

A multi-channel digital technique has been adopted for analysing wave patterns of stresses and strains in structures, particularly under dynamic conditions. This technique provides adequate signal to noise discrimination and high sensitivity for very small (few milli-volts) and slow varying signals (few Hz to 100 Hz.), and A-D conversion accompined by live display during the course of data gathering and computer compatible output. This system also provides fast response because of inherent 50 MHz digitising speed and a large dynamic range of 1024 discrete signal steps. The signals can be suitably fed to the A-D converter (50 MHz) or can be analysed employing frequency modulation techniques and time mode operation of the analyser. The data can be gathered in the field on cassette tapes and replayed in the laboratory for detailed analysis. This technique would provide a versatile system for dynamic analysis of structures under varying conditions. e.g. structures in nuclear power systems, such as testing of end fittings, calandria, vibration testing and measurements exploying pressure transducers.

Dynamic strain analysis of structures employing digital signal processing, storage and display

International Nuclear Information System (INIS)

Patwardhan, P.K.; Misra, V.M.; Kumar, Surendra

1975-01-01

A multi-channel digital technique has been adopted for analysing wave patterns of stresses and strains in structures, particularly under dynamic conditions. This technique provides adequate signal to noise discrimination and high sensitivity for very small (few milli-volts) and slow varying signals (few Hz to 100 Hz.), A-D conversion accompined by live display during the course of data gathering and computer compatible output. This system also provides fast response because of inherent 50 MHz digitising speed and a large dynamic range of 1024 discrete signal steps. The signals can be suitably fed to the A-D converter (50 MHz) or can be analysed employing frequency modulation techniques and time mode operation of the analyser. The data can be gathered in the field on cassette tapes and replayed in the laboratory for detailed analysis. This technique would provide a versatile system for dynamic analysis of structures under varying conditions. e.g. structures in nuclear power systems, such as testing of end fittings, calandria, vibration testing and measurements exploying pressure transducers. (author)

SLARette Mark 2 system

International Nuclear Information System (INIS)

Burnett, D.J.

1992-01-01

The SLAR (Spacer Location and Repositioning) program has developed the technology and tooling necessary to locate and reposition the fuel channel spacers that separate the pressure tube from the calandria tube in a CANDU reactor. The in-channel SLAR tool contains all the inspection probes, and is capable of moving spacers under remote control. The SLAR inspection computer system translates all eddy currents and ultrasonic signals from the in-channel tool into various graphic displays. The in-channel SLAR tool can be delivered and manipulated in a fuel channel by either a SLAR delivery machine or a SLARette delivery machine. The SLAR delivery machine consists of a modified fuelling machine, and is capable of operating under totally remote control in automatic or semi-automatic mode. The SLARette delivery machine is a smaller less automated version, which was designed to be quickly installed, operated, and removed from a limited number of fuel channels during regular annual maintenance outages. This paper describes the design and operation of the SLARette Mark 2 system. 5 figs

Health physics experiences in achieving ALARA exposures to plant personnel at NAPS

International Nuclear Information System (INIS)

Ramakrishna, V.; Lal Chand

2000-01-01

Unit 1 of NAPS achieved first criticality on 12.3.1989 and Unit 2 achieved on 24.10.1991. Till the end of Feb-2000 these units have completed 1890 and 1811 full power days respectively. The performance of NAPS was expected to be better than the earlier Indian reactors in respect of safe production as well as cumulative radiation exposures. This is because of the major design improvements like: fully double containment system, elimination of 41 Ar by introducing light water in calandria vault, reduction of core based fuel failure rate, separation of high radiation equipment to no occupancy areas during normal operation, a separate purification building for the purification of both moderator and PHT systems, a better layout of equipment and plant areas, elimination of unnecessary equipment in various systems besides ensuring the reliability of equipment for safe operation, selection of materials with low corrosion and activation characteristics etc. In this paper, the operational health physics experiences at NAPS to achieve ALARA exposures to plant personnel are described briefly. (author)

Implementation of Moderator Circulation Test Temperature Measurement System

International Nuclear Information System (INIS)

Lim, Yeong Muk; Hong, Seok Boong; Kim, Min Seok; Choi, Hwa Rim; Kim, Hyung Shin

2016-01-01

Moderator Circulation Test(MCT) facility is 1/4 scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. MCT is an equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV(Particle Image Velocimetry) is performed under the iso-thermal test conditions. The Korea Atomic Energy Research Institute (KAERI) started implementation of MCT Temperature Measurement System (TMS) using multiple infrared sensors. To control multiple infrared sensors, MCT TMS is implemented using National Instruments (NI) LabVIEW programming language. The MCT TMS is implemented to measure sensor data of multiple infrared sensors using the LabVIEW. The 35 sensor pipes of MCT TMS are divided into 2 ports to meet the minimum measurement time of 0.2 seconds. The software of MCT TMS is designed using collection function and processing function. The MCT TMS has the function of monitoring the states of multiple infrared sensors. The GUI screen of MCT TMS is composed of sensor pipe categories for user

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

Energy Technology Data Exchange (ETDEWEB)

Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I

1996-07-01

In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA1 and Embalse Argentine NPP

International Nuclear Information System (INIS)

Lescano, V.H.; Wentzeis, L.M.; Guevara, M.; Moreno, C.; Pineyro, J.

1996-01-01

In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

Conceptual design of a quasi-homogeneous pressurized heavy water reactor to be operated in the closed Th-U233 fuel cycle

International Nuclear Information System (INIS)

1979-06-01

This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed

Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

International Nuclear Information System (INIS)

Chang, S.M.; Kim, H.T.

2013-01-01

CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

Specific aspects for Cernavoda - Unit 1 NPP life assurance

International Nuclear Information System (INIS)

Rucareanu, R.

2002-01-01

Full text: The main scope of a Plant Life Management Program is to operate the NPP in a safe manner and at a competitive cost during the reactor life. To achieve this goal, it is important to continuously evaluate the degradation of the main structures and components of the NPP. Background -Cernavoda NPP design life is 30 years. Compared with this target, the operation history is not long (Unit 1 is in commercial operation since 1997). It is still important to begin a plant life management program early to identify the critical components and structures, to establish the data needed for their monitoring and to find methods to mitigate their degradation. A specific aspect for Cernavoda NPP - Unit 1 is the long delay between the fabrication of the main components and the start-up. Most components were procured 10-15 years before start-up. First criticality was achieved in 1996, but the containment perimeter wall sliding was complete in 1983, the Calandria vessel was installed in 1985, the Steam Generators were in position in 1987, the fuel channels were installed in 1989. In evaluating the history of these components, the preservation period must be observed. For Unit 2, which will be in service around 2005, the delay will be longer. For this reason, CNCAN (the Romanian Regulatory Authority) imposed, as a condition to resume the work, to evaluate the ageing of the existing components and structures in order to establish their acceptability for use in the plant. The results of this evaluation can be used as references for subsequent evaluations. Plant Life Assurance Programme - The first step of a PLIM programme is to identify the components and structures that are important for the plant life management. Critical components and structures selection is done using the following criteria: safety criteria - components and structures whose failure can cause a release of radioactivity or which have to mitigate the release of radioactivity in case of a failure of other

Plant condition assessments as a requirement before major investment in life extension for a CANDU nuclear power plant

International Nuclear Information System (INIS)

Aubray, Marc

2002-01-01

Full text: Since, to extend the life of a CANDU-6 reactor beyond its original design life requires the replacement of reactor components (380 pressure and calandria tubes), a major investment will have to be done. After a preliminary technical and economical feasibility study, Hydro- Quebec, owner of the Gentilly-2 NPP, has decided to perform a more detailed assessment to: 1. Get assurance that it is technically and economically viable to extend Gentilly-2 for another 20 years beyond the original design life; 2. Identify the detailed work to be done during the refurbishment period planned in 2008-2009; 3. Define the overall cost and the general schedule of the refurbishment phase; 4. Ensure an adequate licensing strategy to restart after refurbishment; 5. Complete all the Environmental Impact Studies required to obtain the government authorizations. The business case to support the refurbishment of Gentilly-2 has to take in consideration the reactor core components, which will be the major work to be completed during refurbishment. In summary the following main component will have to be changed or refreshed: The pressure and calandria tubes and the feeders (partial replacement only) (ageing mechanisms); The control computers (obsolescence); The condenser tubes (tubes plugging); The turbine control and electric-governor (obsolescence). An extensive campaign is under way to assess the 'health' of the station systems, structures and components (SSC). Two processes have been used for this assessment: Plant Life Management Studies (PLIM) for approximately 10 critical SSC or families of SSC (PLIM Studies); Condition Assessment Studies for other SSC with a lower impact on the Plant production or safety). The PLIM Studies are done on SSC's, which were judged critical because they are not replaceable (Reactor Building, Calandria), or that their failure could have a significant impact on safety or production (electrical motors, majors pumps, heat exchangers and pressure

Severe Accident R and D for Enhanced CANDU-6 Reactors

International Nuclear Information System (INIS)

Nitheanandan, Thambiayah

2012-01-01

CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

DHRUVA

International Nuclear Information System (INIS)

1984-01-01

The pictorial brochure describes DHRUVA reactor which is the fifth research reactor built by the Bhabha Atomic Research Reactor, Bombay. It is a 100 MWt natural uranium fuelled, heavy water moderated and cooled thermal reactor. Salient design data of the reactor building, reactor block and major reactor components are given. Salient features of the reactor core (including fuel assemblies, coolant channels, reactor vessel, calandria), heat transport system, reactor safety system, fuel handling system and radioactive waste management are described in brief. The maximal thermal neutron flux of 1.8x10 14 neutrons per cm 2 per second will be available. Facilities provided for research are two engineering loops, a number of horizontal, radial and tangetial tubes, microprocessor controlled neutron spectrometers, a pneumatic carrier facility for short time irradiation of samples, and systems to maintain the samples at different conditions. Two engineering loops will be used for investigating performance of prototype fuel elements and materials under simulated operating conditions in a power reactor. The reactor will also produce radioisotopes of high specific activity. Major components of the reactor such as reactor vessel, fuelling machines, heat exchangers, coolant circulating pumps have been fabricated indigeneously. (M.G.B.)

MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Mathew, P.M

2001-10-01

To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

Severe accident analysis of a steam generator tube rupture accident using MAAP-CANDU to support level 2 PSA for the Point Lepreau Generating Station Refurbishment Project

Energy Technology Data Exchange (ETDEWEB)

Petoukhov, S.M.; Brown, M.J. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

2015-07-01

A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP-CANDU code was used to simulate the progression of postulated severe core damage accidents and fission product releases. This paper discusses the results for the reference case of the Steam Generator Tube Rupture initiating event. The reference case, dictated by the Level 2 Probabilistic Safety Assessment, was extreme and assumed most safety-related plant systems were not available: all steam generator feedwater; the emergency water supply; the moderator, shield and shutdown cooling systems; and all stages of emergency core cooling. The reference case also did not credit any post Fukushima lessons or any emergency mitigating equipment. The reference simulation predicted severe core damage beginning at 3.7 h, containment failure at 6.4 h, moderator boil off by 8.2 h, and calandria vessel failure at 42 h. A total release of 5.3% of the initial inventory of radioactive isotopes of Cs, Rb and I was predicted by the end of the simulation (139 h). Almost all noble gas fission products were released to the environment, primarily after the containment failure. No hydrogen/carbon monoxide burning was predicted. (author)

Analysis of the fast-neutron spectrum inside the experimental cavity of the NRU Mk4 FN rod

International Nuclear Information System (INIS)

Leung, T.C.

1995-01-01

The fast-neutron (FN) rods in the NRU reactor provide a facility to study the effects of irradiation on CANDU reactor materials. The Mark 4 (Mk4) FN rods use natural uranium and supply fast-neutrons for experiments on irradiation creep and growth, and corrosion, for pressure- and calandria-tube materials. The neutron fluxes above 1 MeV are up to 2.7x10 17 n.m -2 .s -1 . This paper describes a calculation of the fast-neutron spectrum inside the NRU Mk4 FN rod cavity. The calculation was performed using the WIMS-AECL code, which is a multi-group transport code with two dimensional capabilities using the collision-probability method. Results for the fast-neutron spectrum above 1 MeV are presented in nine groups. The analysis confirms that the spectrum in the fast-neutron irradiation facility in NRU is representative of the actual irradiation spectrum for fast-neutron damage in a CANDU reactor. The effects of changes in specimen holder size, temperature, coolant density and fuel burnup on the fast neutron spectrum are also presented. (author). 9 refs., 3 tabs., 4 figs

A layman's guide to radiation-induced deformation processes in zirconium alloys

International Nuclear Information System (INIS)

Dutton, R.

1990-07-01

The fuel channel (comprising a pressure tube and a calandria tube fabricated from zirconium alloys) in a CANDU reactor undergoes shape changes because of radiation-induced deformation. This is a consequence of the microstructural modification arising from radiation damage produced by the fast-neutron flux. This report summarizes our current understanding of the physical processes responsible for the deformation. With the non-specialist reader in mind, the underlying mechanisms are described in a manner that avoids much of the associated technical terminology. Thus, the basic concepts of plasticity in a crystalline material are introduced and related to the various microstructural defects created during irradiation. In particular, the mechanisms of creep (a time-dependent strain activated by an applied stress) and growth (a time-dependent strain occurring in the absence of stress) are discussed in a non-technical language assisted by simple diagrams. Reference is made to both theoretical investigations (avoiding mathematical complexity) and experimental measurements. It is shown how the qualitative and quantitative knowledge can be used to derive a predictive model for reactor designers and operators. The current status of such a model is evaluated and suggestions for future improvements made

Numerical Analysis of CANDU-6 Moderator System Using OpenFOAM

International Nuclear Information System (INIS)

Chang, Se Myong; Kim, Hyoung Tae

2012-01-01

On the moderator of CANDU-6 reactor, thanks to the rapid development of CFD (Computational Fluid Dynamics), the 1-D model code can be substituted to the 3-D simulation codes. The three-dimensional computation becomes not so expensive that now we can enjoy the benefit of innovation about CFD technology. In this study, we have modeled the Calandria tank system as simplified models preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors. The use of OpenFOAM is a very important point for the present study. The OpenFOAM is based on the object-oriented programming using C++ language. The solvers and libraries of physical properties, for example, are declared as classes to produce a new code with the reproduction from the existing classes. As this code is fully open to the public, the development of CFD code with OpenFOAM should be very prospective to the future design of system codes, not just restricted in the area of hydro-thermal system concerning atomic reactors

low dose irradiation growth in zirconium

International Nuclear Information System (INIS)

Fortis, A.M.

1987-01-01

Low dose neutron irradiation growth in textured and recrystallized zirconium, is studied, at the Candu Reactors Calandria temperature (340 K) and at 77 K. It was necessary to design and build 1: A facility to irradiate at high temperatures, which was installed in the Argentine Atomic Energy Commission's RA1 Reactor; 2: Devices to carry out thermal recoveries, and 3: Devices for 'in situ' measurements of dimensional changes. The first growth kinetics curves were obtained at 365 K and at 77 K in a cryostat under neutron fluxes of similar spectra. Irradiation growth experiments were made in zirconium doped with fissionable material (0,1 at % 235 U). In this way an equivalent dose two orders of magnitude greater than the reactor's fast neutrons dose was obtained, significantly reducing the irradiation time. The specimens used were bimetallic couples, thus obtaining a great accuracy in the measurements. The results allow to determine that the dislocation loops are the main cause of irradiation growth in recrystallized zirconium. Furthermore, it is shown the importance of 'in situ' measurements as a way to avoid the effect that temperature changes have in the final growth measurement; since they can modify the residual stresses and the overconcentrations of defects. (M.E.L.) [es

Options for management of containment integrity during severe accident in Indian PHWR

International Nuclear Information System (INIS)

Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, N.; Nair, Suma R.

2015-01-01

Severe accident progressions have the potential to raise the containment pressure beyond the design pressure of the structure. Although the load withstanding capability of the containment structure has been assessed to be substantially higher than the design pressure of the structure (typically 2 times of design pressure), it is possible that a few components of Containment System may degrade leading to excessive release of radioactive fission gases at ground level. Additionally, possible cracks in the concrete of the containment at high pressure may aggravate the release at ground level. Over and above, maintaining high containment pressure high for a longer period increases the ground level release due to leakage from the containment, which effect on dose might be high. For maintaining the Integrity of the Containment, containment pressure can be reduced by either energy management system such as removing the heat from the calandria vault (CVWC) water by using CV water heat exchanger intermittently or reliving the containment atmosphere either through Primary Containment Controlled Discharge (PCCD) or Containment Filtered Venting System (CFVS). Further, it is necessary that these provisions must be initiated below design pressure. This paper presents the analysis for the containment depressurization by using CVWC system restored, manual opening of (PCCD) line and operation of CFVS during the progressions of the accident

Model based optimization of driver-pickup separation for eddy current measurement of gap

Science.gov (United States)

Klein, G.; Morelli, J.; Krause, T. W.

2018-04-01

The fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of a pressure tube (PT) contained within a larger diameter calandria tube (CT). The separation between the tubes, known as the PT-CT gap, ensures PT hydride blisters, which could lead to potential cracking of the PT, do not develop. Therefore, accurate measurements are required to confirm that contact between PT and CT is not imminent. Gap measurement uses an eddy current probe. However this probe is sensitive to lift-off variations, which can adversely affect estimated gap. A validated analytical flat plate model of eddy current response to gap was used to examine the effect of driver-pickup spacing on lift-off and response to gap at a frequency of 4 kHz, which is used for in-reactor measurements. This model was compared against, and shown to have good agreement with, a COMSOL® finite element method (FEM) model. The optimum coil separation, which included the constraint of coil size, was found to be 11 mm, resulting in a phase response between lift-off and response to change in gap of 66°. This work demonstrates the advantages of using analytical models for optimizing coil designs for measurement of parameters that may negatively influence the outcome of an inspection measurement.

Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

International Nuclear Information System (INIS)

Kim, Young Suk

2007-01-01

In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

Development of demonstration advanced thermal reactor

Energy Technology Data Exchange (ETDEWEB)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

1982-08-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

Development of microstructure in thermomechanical processing of zirconium alloys

International Nuclear Information System (INIS)

Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

Parametric studies to establish natural circulation in advanced heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Bhatia, S K; Dhawan, M L [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

Design of Advanced Heavy Water Reactor (AHWR) is in progress. It consists of vertical pressure tubes with boiling light water coolant flowing through the tubes and heavy water moderator in the calandria. In PHWRs, core heat removal is through forced circulation of the coolant by PHT pumps. In AHWR, no PHT pumps are used and core heat is carried away by natural circulation of the coolant due to density difference between steam/water mixture inside the core and the water region outside the core. This passive means of core heat removal results in a number of benefits viz. (a) extra length of piping, valves, instruments, power supply and control systems for functioning of instruments are eliminated, (b) plant layout is simplified, (c) maintenance of valves and instruments is reduced. Natural circulation in AHWR is achieved by keeping the steam drum at a sufficient height above the core to get the required driving force. The loop height depends on many factors i.e. channel power, V{sub c}/V{sub f} ratio (ratio of coolant volume to fuel volume) and core height. The effect of these parameters on the loop height to establish natural circulation have been studied and presented. (author). 1 ref., 1 fig., 1 tab.

Repairs in 104 Gy/h?

International Nuclear Information System (INIS)

Anon.

1994-01-01

In 1989 it was found that in each unit of the MAPS Candu nuclear power station in India the centre portion of the heavy water inlet manifold opposite the 300 mm inlet pipe had torn away. Equipment for remote inspection, repair and removal of debris in an area with constricted and difficult access and radiation fields of about 10 4 Gy/h was developed by Ricardo Hitec of the United Kingdom. This consists of a work performing manipulator, TV viewing systems and a posting tube manipulator for insertion of tools and debris containers into the calandria. A variety of special end effector and tools were also developed jointly with AEA Technology. A first repair campaign was carried out on Unit 1 in 1991. Following a detailed TV survey of the damage a reappraisal of the situation was undertaken and a programme of equipment enhancement carried out. In July 1992 a second repair campaign took place on Unit 2. The difficulties encountered and the degree of success achieved are described. Work proceeded at an intensive level for 14 days when the campaign was ended in view of the exhaustion both of personnel and the equipment. Although more work could have been done a major improvement had been achieved. (UK)

NRU vessel repair and return to service: enhancing a Canadian R and D asset for the future

Energy Technology Data Exchange (ETDEWEB)

Cox, D.S.; Arnold, J.B.; Lee, J.K., E-mail: coxd@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

The National Research Universal (NRU) reactor was successfully returned to high power operation in 2010 August, after completing extensive inspections and repairs of the calandria vessel, in response to a small leak of heavy water that was discovered in 2009 May. The aluminum alloy vessel material had corroded from the outside surface over many years, and required application of weld build-up and plated weld repair at ten locations, executed by remote tooling deployed from inside the reactor vessel. Many specialized remotely operated tools were developed for inspection, sampling and repair operations. In parallel with the repair efforts, important maintenance activities were performed, enabled by the de-fuelled state with the heavy water drained from the vessel. Two annual inspection cycles have now been completed since the reactor was returned to high power operation, confirming the fitness of the vessel for continued operation. The NRU reactor is now entering its 56th year of operation and the current operating licence extends to 2016. Based on the outcome of a comprehensive Integrated Safety Review, AECL is continuing to implement major equipment upgrades and process improvements to support safe and reliable operation through 2021, for the benefit of Canadians and the world. (author)

Prediction of the Inlet Nozzle Velocity Profiles for the CANDU-6 Moderator Analysis

International Nuclear Information System (INIS)

Yoon, Churl; Park, Joo Hwan

2006-01-01

For the moderator analysis of the CANDU reactors in Korea, predicting local moderator subcooling in the Calandria vessels is one of the main concerns for the estimation of heat sink capability of moderator under LOCA transients. The moderator circulation pattern is determined by the combined forces of the inlet jet momentum and the buoyancy flow. Even though the inlet boundary condition plays an important role in determining the moderator circulations, no measured data of detailed inlet velocity profiles is available. The purpose of this study is to produce the velocity profiles at the inlet nozzles by a CFD simulation. To produce the velocity vector fields at the inlet nozzle surfaces, the internal flows in the nozzle assembly were simulated by using a commercial CFD code, CFX-5.7. In the reference, the analytical capability of CFX-5.7 had been estimated by a validation of the CFD code against available experimental data for separate flow phenomena. Various turbulence models and grid spacing had been also tested. In the following section, the interface treatment between the computational domains would be explained. In section 3, the inlet nozzle flow through the CANDU moderator nozzle assembly was predicted by using the obtained technology of the CFD simulation

Development of demonstration advanced thermal reactor

International Nuclear Information System (INIS)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

1982-01-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

Multiphysical Simulation of PT-CT Contact with Outer Boundary Condition

Energy Technology Data Exchange (ETDEWEB)

Chang, Se-Myong [Kunsan National Univ., Gunsan (Korea, Republic of); Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The present study is about preliminary calculation results for these ICSP activity works, where the COMSOL Multiphysics code is used to simulate plastic deformation of a pressure tube as a result of the interaction of stress and temperature. It is shown that the thermal stress model of COMSOL is compatible to simulate the multiple heat transfers (including the radiation heat transfer and heat conduction) and stress strain in the simplified 2-D problem. The benchmark test result for radiation heat transfer is in good agreement with the analytical solution for the concentric configuration of PT(pressure tube) and CT(calandria tube). In this paper, the authors did an open computation of these multi-physical phenomena by changing the outer boundary condition of CT according to the experimental result of ICSP. A series of simulation has been done based on the benchmark test proposed by IAEA/ICSP. The unsteady multi-physics was treated some numerical models with COMSOL. The comparison with CATHENA code is verified as a good agreement as we increase the accuracy of numerical method, Gaussian quadrature. The open computation for the validation of this numerical code is still on-going, and the temperature inside and outside the PT shows a very good agreement.

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Point Lepreau refurbishment: plant condition assessment

International Nuclear Information System (INIS)

Allen, P.J.; Soulard, M.R.; David, F.; Clefton, G.; Weeks, R.

2001-01-01

New Brunswick Power (NB Power) has initiated a study into the refurbishment of the Point Lepreau Generating Station, with the objective to extend plant operation another 25 to 30 years. The end product of this study will be a business case that compares the costs of refurbishing Point Lepreau with costs of alternate means of generation. The Project Execution Plan and business case are being developed by an integrated team of AECL, NB Power and subcontractor staff under the project management of AECL. The refurbishment scope will include replacement of the pressure tubes, calandria tubes and part of the feeder piping. Planning of these replacements is part of the refurbishment study work. Planning is also underway for the environmental, safety and licensing issues that would need to be addressed to ensure future operation of the unit. In addition to these studies, a systematic review of the plant has been carried out to determine what other equipment refurbishment or replacement will be required due to ageing or obsolescence of plant equipment. This Plant Condition Assessment (PCA) follows a highly structured approach to ensure consistency. This paper presents an overview of the engineering process and the main findings from the work. (author)

Multiphysical Simulation of PT-CT Contact with Outer Boundary Condition

International Nuclear Information System (INIS)

Chang, Se-Myong; Kim, Hyoung Tae

2016-01-01

The present study is about preliminary calculation results for these ICSP activity works, where the COMSOL Multiphysics code is used to simulate plastic deformation of a pressure tube as a result of the interaction of stress and temperature. It is shown that the thermal stress model of COMSOL is compatible to simulate the multiple heat transfers (including the radiation heat transfer and heat conduction) and stress strain in the simplified 2-D problem. The benchmark test result for radiation heat transfer is in good agreement with the analytical solution for the concentric configuration of PT(pressure tube) and CT(calandria tube). In this paper, the authors did an open computation of these multi-physical phenomena by changing the outer boundary condition of CT according to the experimental result of ICSP. A series of simulation has been done based on the benchmark test proposed by IAEA/ICSP. The unsteady multi-physics was treated some numerical models with COMSOL. The comparison with CATHENA code is verified as a good agreement as we increase the accuracy of numerical method, Gaussian quadrature. The open computation for the validation of this numerical code is still on-going, and the temperature inside and outside the PT shows a very good agreement

Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2004-02-15

The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

CFD Analysis for the Steady State Test of CS28-1 Simulating High Temperature Chemical Reactions in CANDU Fuel Channel

International Nuclear Information System (INIS)

Park, Ju Hwan; Kang, Hyung Seok; Rhee, Bo Wook

2006-05-01

The establishment of safety analysis system and technology for CANDU reactors has been performed at KAERI. As for one of these researches, single CANDU fuel bundle has been simulated by CATHENA for the post-blowdown event to consider the complicated geometry and heat transfer in the fuel channel. In the previous LBLOCA analysis methodology adopted for Wolsong 2, 3, 4 licensing, the fuel channel blowdown phase was analyzed by a CANDU system analysis code CATHENA and the post-blowdown phase of fuel channel was analyzed by CHAN-IIA code. The use of one computer code in consecutive analyses appeared to be desirable for consistency and simplicity in the safety analysis process. However, validation of the high temperature post-blowdown fuel channel model in the CATHENA before being used in the accident analysis is necessary. Experimental data for the 37-element fuel bundle that fueled CANDU-6 has not been performed. The benchmark problems for the 37-element fuel bundle using CFD code will be compared with the test results of the 28-element fuel bundle in the CS28-1 experiment. A full grid model of FES to the calandria tube simulating the test section was generated. The number of the generated mesh in the grid model was 4,324,340 cells. The boundary and heat source conditions, and properties data in the CFD analysis were given according to the test results and reference data. Thermal hydraulic phenomena in the fuel channel were simulated by a compressible flow, a highly turbulent flow, and a convection/conduction/radiation heat transfer. The natural convection flow of CO 2 due to a large temperature difference in the gap between the pressure and the calandria tubes was treated by Boussinesq's buoyancy model. The CFD results showed good agreement with the test results as a whole. The inner/middle/outer FES temperature distributions of the CFD results showed a small overestimated value of about 30 .deg. C at the entrance region, but good agreement at the outlet region. The

Experimental investigation of quench and re-wetting temperatures of hot horizontal tubes well above the limiting temperature for solid–liquid contact

Energy Technology Data Exchange (ETDEWEB)

Takrouri, Kifah, E-mail: takroukj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Luxat, John, E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Hamed, Mohamed [Thermal Processing Laboratory (TPL), Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada)

2017-01-15

Highlights: • Quench and re-wetting temperatures were measured upon jet quenching of hot cylindrical tubes. • Correlations have been developed and provided good fit of data. • Quench and re-wetting temperatures were found to greatly depend on water subcooling. • Stagnation point showed higher quench and re-wetting temperatures than other locations. • Quench temperature decreased by increasing surface curvature and tube conductivity. • Re-wetting temperature is a weak function of both variables. - Abstract: Quench cooling of a hot dry surface involves the rapid decrease in surface temperature resulting from bringing the hot surface into sudden contact with a coolant at a lower temperature. Quench temperature is the onset of the rapid decrease in surface temperature and corresponds to the onset of destabilization of a vapor film that exists between the hot surface and the coolant. Situations involving quench cooling are encountered in a number of postulated accidents in Canada Deuterium Uranium CANDU reactors, such as the quench of a hot calandria tube in certain Loss of Coolant Accidents LOCA. If the calandria tube temperature is not reduced by initiation of quench heat transfer, then this may lead to subsequent fuel channel failure and for this accident knowledge of quench heat transfer characteristics is of great importance. In this study, a Water Quench Facility WQF has been designed and built at the Thermal Processing Laboratory TPL at McMaster University and a series of experimental tests were carried out to investigate the quench of hot horizontal tubes using a vertical rectangular water multi-jet system. The tubes were heated to a temperature between 380 and 780 °C then cooled to the jet temperature. The temperature variation with time in tube circumferential and axial directions was measured. The two-phase flow behavior and the propagation of the re-wetting front around and along the tubes were simultaneously observed using a high-speed camera

Simulation and analysis of the thermal and deformation behaviour of `as-received` and `hydrided` pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

Energy Technology Data Exchange (ETDEWEB)

Muir, W C; Bayoumi, M H [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 {mu}g/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs.

Advanced CANDU Design With Negative Power Feedback

International Nuclear Information System (INIS)

Andang-Widi-Harto; Muslim

2004-01-01

The problem of positive power feedback in the recent PHWR-CANDU design, especially related to coolant void increase, will be overcame by the use of dual moderator concept, in which two moderator systems are used, i.e. a main moderator outside the calandria tube and an annular moderator inside the annular space. Annular moderator is allowed to boil in the case of overheating. The numerical calculations have been performed for two core design namely HWR-DM-ST and HWR-DM-XI which can reach burn up of 16,000 and 17,500 MWd/ ton U respectively. The results for the two designs is that the values of k at fully annular moderator filling condition are 1.0054 (HWR-DM-ST) and 1.0019 (HWR-DM-XI), while at completely empty annular moderator condition are 0.9634 (HWR-DM-ST) and 0.9143 (HWR-DM-XI). The decrease of coolant flow rate from 3,043 kg/s to 853 kg/s decrease k values of 0.0109 (HWR-DM-ST) and 0.0232 (HWR-DM-XI). While increasing inlet coolant enthalpy from 2,950 kJ/kg to 3,175 kJ/kg decreases of k values of 0.0074 (HWR-DM-ST) and 0.0239 (HWR-DM-XI). Thus, it can be summarized that the HWR-DM design has negative power reactivity feedback.(author)

Conceptual design of a pressure tube light water reactor with variable moderator control

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2012-01-01

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

Analytical tools and methodologies for evaluation of residual life of contacting pressure tubes in the early generation of Indian PHWRs

International Nuclear Information System (INIS)

Sinha, S.K.; Madhusoodanan, K.; Rupani, B.B.; Sinha, R.K.

2002-01-01

In-service life of a contacting Zircaloy-2 pressure tube (PT) in the earlier generation of Indian PHWRs, is limited mainly due to the accelerated hydrogen pick-up and nucleation and growth of hydride blister(s) at the cold spot(s) formed on outside surface of pressure tube as a result of its contact with the calandria tube (CT). The activities involving development of the analytical models for simulating the degradation mechanisms leading to PT-CT contact and the methodologies for the revaluation of their safe life under such condition form the important part of our extensive programme for the life management of contacting pressure tubes. Since after the PT-CT contact, rate of hydrogen pick-up and nucleation and growth of hydride blisters govern the safe residual life of the pressure tube, two analytical models (a) hydrogen pick-up model ('HYCON') and (b) model for the nucleation and growth of hydride blister at the contact spot ('BLIST -2D') have been developed in-house to estimate the extent of degradation caused by them. Along with them, a methodology for evaluation of safe residual life has also been formulated for evaluating the safe residual life of the contacting channels. This paper gives the brief description of the models and the methodologies relevant for the contacting Zircaloy-2 pressure tubes. (author)

In-core LOCA-s: analytical solution for the delayed mixing model for moderator poison concentration

International Nuclear Information System (INIS)

Firla, A.P.

1995-01-01

Solutions to dynamic moderator poison concentration model with delayed mixing under single pressure tube / calandria tube rupture scenario are discussed. Such a model is described by a delay differential equation, and for such equations the standard ways of solution are not directly applicable. In the paper an exact, direct time-domain analytical solution to the delayed mixing model is presented and discussed. The obtained solution has a 'marching' form and is easy to calculate numerically. Results of the numerical calculations based on the analytical solution indicate that for the expected range of mixing times the existing uniform mixing model is a good representation of the moderator poison mixing process for single PT/CT breaks. However, for postulated multi-pipe breaks ( which is very unlikely to occur ) the uniform mixing model is not adequate any more; at the same time an 'approximate' solution based on Laplace transform significantly overpredicts the rate of poison concentration decrease, resulting in excessive increase in the moderator dilution factor. In this situation the true, analytical solution must be used. The analytical solution presented in the paper may also serve as a bench-mark test for the accuracy of the existing poison mixing models. Moreover, because of the existing oscillatory tendency of the solution, special care must be taken in using delay differential models in other applications. (author). 3 refs., 3 tabs., 8 figs

Simulation and analysis of the thermal and deformation behaviour of 'as-received' and 'hydrided' pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

International Nuclear Information System (INIS)

Muir, W.C.; Bayoumi, M.H.

1995-01-01

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 μg/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs

Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

International Nuclear Information System (INIS)

Talpariu, C; Talpariu, J.; Matei, C.

2001-01-01

Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities

Progress report, Chemistry and Materials Division, 1 April - 30 June, 1981

International Nuclear Information System (INIS)

1981-08-01

The work of the Division in the areas of solid state science, radiation, physical and analytical chemistry, and materials science during the quarter is described. Measurements of ion stopping power have emphasized the importance of axial symmetry and may be used to show the contribution of nuclear inelastic events to stopping processes. Enhancement of ion scattering at 180 degrees can occur even in the first few layers of a single crystal of gold implanted with heavy atoms. Agreement has been obtained between experimental and calculated rates for dechanneling of protons in gold. The rate of decomposition of HOI in aqueous solutions has been determined. The effects of radiation on dithiothreitol is being studied. Laser photochemistry work includes investigations of multiphoton dissociation and of laser-induced zirconium isotope separation. A method has been found for the preparation of oxygen gas samples for the determination of oxygen isotope ratios in water, and high-performance liquid chromatography has been applied to metals in ground water. Sputtered coatings of stainless steel on the surface of zircaloy fuel cladding reduce the oxidation rate in steam. A theoretically-based design equation for irradiation growth of pressure tubes has been developed. Studies on the effect of small strains on zircaloy-2 tubing show the need to avoid even small amounts of compressive deformation of calandria tubes

Effect of grain shape and texture on equi-biaxial creep of stress relieved and recrystallized Zircaloy-4

International Nuclear Information System (INIS)

Murty, K.L.; Tanikella, B.V.; Earthman, J.C.

1994-01-01

Zirconium alloys are extensively used in various types of fission reactors both light and heavy water types for different applications, examples being thin-walled tubing to clad radioactive fuel, grids, channels in boiling water reactors (BWRs) as well as pressure and calandria tubes in pressurized heavy water reactors (PHWRs). Biaxial creep behaviors of stress relieved and recrystallized thin-walled tubing of Zircaloy-4 are considered under equal hoop and axial stresses by internal pressurization superimposed with axial load. Both hoop and axial strains were monitored and the ratio of the strain rates along the hoop to axial directions is considered to represent the degree of anisotropy. The slightly stronger hoop direction of the recrystallized material became weaker compared to the axial direction following cold work and a stress-relief anneal. Crystallographic texture was considered in terms of x-ray pole figures from which the crystallite orientation distribution functions (CODF) were derived. A crystal plasticity model based on slip on representative systems was combined with the CODF to predict the creep anisotropy. It was found that the textural differences between the recrystallized and stress-relieved material is believed to invoke anisotropic grain boundary sliding leading to stress enhancement in the hoop direction. This stress enhancement is shown to account for the observed differences in creep behavior between the present equiaxed and columnar grain structures

Development of high pressure conductivity probe (HPCP) for secondary shut down system (SDS-2) of 500 MWe PHWR

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.; Mohan, L.R.

2003-09-01

The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface. This interface moves towards the calandria because of molecular diffusion, temperature difference and physical disturbances in the moderator level. It is proposed to install two numbers of high pressure conductivity probes (HPCP) to monitor the interface movement as well as to provide the safe annunciation value for interface location. On actuation of the SDS-2 signal, high-pressure helium will inject the poison into the moderator to shutdown the reactor. During poison injection, these probes will experience high pressure of nearly 85 kg/sq.cm. Global market survey indicated that conductivity probes having built in temperature sensor are available for a maximum pressure rating of 35 kg/sq.cm. Hence in order to meet the process requirement of SDS-2, the development of HPCP suitable for a pressure of 85 kg/sq.cm. was taken up. Two numbers of such probes were successfully designed, fabricated and evaluated for their performance. The developed conductivity probes fully meet the laid design and performance criteria. The aforesaid development work was a successful endeavour towards indigenisation of high-pressure conductivity probe for future applications. This report deals with the design aspects, fabrication technique, material and performance evajuation criteria and test results of HPCP. (author)

Effects of the plastic deformation and thermal cycles on the mechanical properties of fully recrystallized Zircaloy-4

International Nuclear Information System (INIS)

Litvack, Nicolas

2005-01-01

The development of crystallographic texture in a product depends, for a given material, of its fabrication history. In our case, the evolution of that texture results from a combination of cold working and thermal cycles applied together or separately. In the present work, cold working levels ranging from 50 % to approximately 90 % and different heat treatment cycles has been applied to Zircaloy-4 sheets and tubes. Using X-ray diffraction techniques and the direct pole figure method, the evolution of crystallographic texture has been analyzed for each fabrication route. We observed that cold working levels up to 90 % without intermediate annealing heat treatment do not change significantly the classic angle between basal pole and the normal/radial direction of the product (φ ≅ ± 25 degrees). Furthermore, the application of intermediate cold working levels (50 % - 60 %) and more than two intermediate annealing heat treatments exhibits a marked modification of the basal pole orientation. The basal poles appear now parallel to the normal direction (φ ≅ 0 degrees) of the product. Additionally, the crystallographic texture change observed with X-ray procedures was evaluated by the measure of anisotropic parameters R and P. The results here obtained will be use in the future as a basis for the design of a fabrication route capable to obtain in a HPTR process, seamless calandria tubes strengthened by crystallographic texture. (author) [es

Reactivity initiated accidents and loss of shutdown - 20 years later

International Nuclear Information System (INIS)

Luxat, J.C.

2007-01-01

A review of the safety of Ontario's nuclear power reactors was conducted in 1987 after the Chernobyl accident. As part of this review an analysis was performed of a Loss of Coolant Accident in a Pickering A unit with coincident failure to shutdown. This analysis showed that the power excursion was halted by channel and calandria vessel failures leading to moderator fluid displacement. The containment structure did not fail and, at worst might suffer minor cracking at the top of the dome of the reactor building. Overall the dose consequences of such an accident were no worse than the limiting design basis dual failure event. In the intervening twenty years following this analysis, Significant experimental information has been obtained that relates to power pulse behaviour. This information, together with conservatisms in he original analysis, are reviewed and assessed in this paper. In addition, the issue of reactivity initiated events in other reactor types is reviewed to identify the reactor design characteristics that are of importance in these events. Contrary to popular belief the existence of positive coolant void reactivity is not as significant a factor as it is sometimes stated to be. On balance, with appropriate design measures, no one reactor type can be claimed to be 'more safe' than another. The underlying basis for this statement is articulated in this paper. (author)

Computer Refurbishment

International Nuclear Information System (INIS)

Ichiyen, Norman; Chan, Dominic; Thompson, Paul

2004-01-01

The major activity for the 18-month refurbishment outage at the Point Lepreau Generating Station is the replacement of all 380 fuel channel and calandria tube assemblies and the lower portion of connecting feeder pipes. New Brunswick Power would also take advantage of this outage to conduct a number of repairs, replacements, inspections and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc.). This would allow for an additional 25 to 30 years. Among the systems to be replaced are the PDC's for both shutdown systems. Assessments have been completed for both the SDS1 and SDS2 PDC's, and it has been decided to replace the SDS2 PDCs with the same hardware and software approach that has been used successfully for the Wolsong 2, 3, and 4 and the Qinshan 1 and 2 SDS2 PDCs. For SDS1, it has been decided to use the same software development methodology that was used successfully for the Wolsong and Qinshan called the I A and to use a new hardware platform in order to ensure successful operation for the 25-30 year station operating life. The selected supplier is Triconex, which uses a triple modular redundant architecture that will enhance the robustness/fault tolerance of the design with respect to equipment failures

Analysis of power variation in a CANDU-6 with a loss of moderator

International Nuclear Information System (INIS)

Fan, Y.

2008-01-01

A loss of heavy water in a postulated small failure in the horizontal unpressurized calandria vessel of a CANDU-6 reactor will lead to a drop in the moderator level in the reactor core. The STEPBACK and SETBACK functions at the initial moment of the drop in moderator level ensure a reactor shutdown and a reduction in total reactor power during this 900 seconds postulated transient. If the STEPBACK and SETBACK functions are unavailable, the reactor's regulating system will try to compensate for the negative reactivity resulting from the loss of the moderator. This kind of compensation will lead to power distortions from top to bottom in the reactor core. .Comparisons of different moderator leakage rates were used in the analysis to determine the relationships between the power and the moderator leakage rates. Maximum bundle and channel powers obtained were insensitive to the moderator leakage rate. .In a complete analysis for a moderator leakage rate of 40 1/s, it was found that, without the STEPBACK and SETBACK functions, serious power distortions would occur during the 900 seconds transient. The maximization of bundle and channel power during this transient happened in the bottom part of the reactor , and the regulating system worsened this power distortion. .From the above analysis, it was concluded that the maximum bundle power attained during the loss of the moderator was 1.18% of its initial value. The risk of bundle dryout was, therefore, quite small. (author)

Nuclear power - replacement of pressure tubes in CANDU reactors

International Nuclear Information System (INIS)

Anon.

1992-01-01

The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

CANDU safety under severe accidents

International Nuclear Information System (INIS)

Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

1996-01-01

The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

Three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

International Nuclear Information System (INIS)

Yu, S.-O.; Kim, M.; Kim, H.-J.

2002-01-01

A CANDU reactor has the unique features and the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors. If there is the loss of coolant accident (LOCA) and a coincident failure of the emergency coolant injection (ECI) system, the heavy water moderator is continuously cooled, providing a heat sink for decay heat produced in the fuel. Therefore, it is one of major concerns to estimate the local subcooling of moderator inside the calandria vessel under postulated accident in CANDU safety analyses. The Canadian Nuclear Safety Commission (CNSC), a regulatory body in Canada, categorized the integrity of moderator as a generic safety issue and recommended that a series of experimental works be performed to verify the safety evaluation codes for individual simulated condition of nuclear power plant, comparing with the results of three-dimensional experimental data. In this study, three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL (Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. Three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. (author)

Las aves de distribución mediterránea en el País Vasco: abundancia y tendencia poblacional en el sur de Álava

Directory of Open Access Journals (Sweden)

Gainzarain, J.A., Belamendia, G.

2015-01-01

Full Text Available Varias especies de aves, entre ellas algunas catalogadas como amenazadas, presentan en la Comunidad Autónoma Vasca una distribución meridional centrada en la Rioja Alavesa. Mediante transectos efectuados en esta comarca en las primaveras de 2012 y 2013, se obtuvieron datos sobre la abundancia de estas especies en diferentes tipos de hábitat. Con el fin de conocer la evolución reciente de sus poblaciones, esta información se comparó con la de dos estudios anteriores, llevados a cabo en 1988/89 y 1994. Nuestros datos revelan que el alcaudón real Lanius meridionalis se ha extinguido como nidificante en la comarca y que, junto con esta especie, la tórtola europea Streptopelia turtur, la calandria común Melanocorypha calandra y el escribano hortelano Emberiza hortulana son las que han experimentado una disminución más marcada. En sentido contrario, la curruca cabecinegra Sylvia melanocephala ha colonizado la comarca después de 1988, y la terrera común Calandrella brachydactyla probablemente haya aumentado sus efectivos. Las tendencias observadas coinciden en gran medida con las registradas en el conjunto de España mediante el programa SACRE. La abundancia global del grupo de especies analizadas ha sufrido un descenso significativo en las dos últimas décadas, paralelo a un notable aumento de la superficie ocupada por el viñedo en la comarca en detrimento de otros usos minoritarios del suelo.

CANDU safety under severe accidents

Energy Technology Data Exchange (ETDEWEB)

Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

1996-12-01

The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

1982-83 annual report

International Nuclear Information System (INIS)

1983-01-01

Atomic Energy of Canada Ltd. continued to improve on its operating results and financial position with record net earnings of $22.6 million, an increase of 15 percent. Commercial revenue from exports was 53 percent. Revenue from nuclear supply and services declined to $218.9 million as reactor projects neared completion or were deferred. Radiation equipment and isotope sales increased by 25 percent to $84.3 million. This revenue included the first shipment of a Canadian cancer therapy accelerator, isotope production from the new cyclotron, and radiopharmaceutical products. Pickering Generating Station revenue was $38.1 million, an increase of 28 percent. Heavy water production increased by 21 percent and unit operating costs decreased by 9 percent due to increased productivity. Operating profit from commercial operations of $25.8 million was slightly lower, due to increased product development costs and marketing expenses. Studies were made of the long-term behaviour of pressure and calandria tubes in CANDU reactors. Cesium and iodine released from fuel in a hypothetical accident would be retained within the system to a much greater degree than was previously believed, according to other basic research studies. Comprehensive safeguards systems and equipment developed in collaboration with the IAEA were installed on the four 600MW CANDU units which began operating. Experiments testing a process to extract useful by-products from used CANDU fuel were completed using facilities in Italy. Future developments using by-products in combination with uranium or thorium will ensure CANDU's long-term viability

Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

Energy Technology Data Exchange (ETDEWEB)

LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

2014-07-01

Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

Chemistry control approach of pre commissioning and power operation of primary and auxiliary system of KGS-3 and 4 and trouble shooting made

International Nuclear Information System (INIS)

Bennet Raj, N.; Sahu, B.S.; Kumar, Vineet; Valluri, J.

2008-01-01

KGS (Kaiga Generating Station) 3 and 4 is a 220 MWe pressurized heavy water reactor (PHWR) using heavy water (D 2 O) as moderator and primary heat coolant and the secondary system is light water which is used to make the steam for generating the power. The chemistry control approach made for the successful commissioning and subsequent power operation of the unit is discussed here. The chemistry control is of two parts first part covers the pre commissioning chemistry control and the second part covers the commissioning chemistry control. During commissioning all systems were preserved by proper chemistry control and regular recirculation of system to avoid stagnancy. The major pre commissioning and commissioning chemistry control are depicted below: Pre commissioning chemistry control of primary heat transport (PHT) system and auxiliaries; Pre commissioning chemistry control of moderator system; Primary heat transport system hot conditioning with light water; Commissioning chemistry control of End Shield System (ESC) and Calandria Vault Cooling (CVC) system; Heavy water addition and its chemistry control in moderator system; and Heavy water addition and its chemistry control in PHT system. During power operation dew point in annular gas monitoring system (AGMS) of KGS unit 3 was maintaining in higher side under recirculation. The increase of dew point could be due to ingress of heavy water or light water. A new device was developed to collect condensate and the chemistry of the condensate was checked. The result indicated the ingress of light water. (author)

An innovative method for on-power radiometry of end-shields of nuclear power plants

International Nuclear Information System (INIS)

Kumar, Gaurav; Gupta, Pankaj; Nawal, Shriram; Gautam, Mahesh; Kakkar, Aman Deep; Yadav, Umed

2012-01-01

Every lndian PHWR reactor calandria is sandwiched within a pair of shield on either side. These shields are perpendicular to the coaxial axis of calandria and are called end-shields. These provide shielding from leakage radiation from reactor core in escaping out to Fuelling Machine vault, thereby significantly reducing the dose rates in the vaults. This has got a direct impact on radiation field in accessible areas. By maintaining low dose rates in accessible areas, the individual and collective doses of radiation workers can be effectively controlled well within the stipulated limits. Thus, it is of utmost importance to ensure adequacy of shielding provided by end-shields. In this context, a limited radiometry exercise is executed after filling of end-shields with steel balls and prior to their installation at designated place. This exercise provides limited inputs along the periphery of end-shield due to limited strength of radiation source, its handling provisions and dose constraints to the individual. In order to ascertain an in-depth analysis of shielding adequacy on-power, different methodologies have been adopted and have certain limitation in precisely pinpointing the affected area/location besides limitation on number of locations that can be monitored at a single stretch. To overcome these important anomalies, a computer based setup has been indigenously designed. The setup essentially comprises of a radiation monitor with wide energy, measuring, temperature and humidity range; a custom designed 25 m long compatible cable with suitable connectors; a laptop with additional cooling arrangement; a configurable interfacing software; thermal shielding for the detector and tying/fixing provisions. The radiation monitor after being properly shielded for thermal impacts is installed on the head of Fuelling Machine. It is connected through long cable to a laptop kept at Fuelling Machine service area with due cooling provisions (as temperature in the area will

Gadolinium depletion event in a CANDU® moderator - causes and recovery

Energy Technology Data Exchange (ETDEWEB)

Evans, D.W.; Price, J.; Swami, D.; Fracalanza, E.; Brett, M.E.; Puzzuoli, F.V.; Garg, A. [Ontario Power Generation, Pickering, Ontario (Canada); Herrmann, O.; Rudolph, A. [Kinectrics Inc., Toronto, Ontario (Canada); Stuart, C.; Glowa, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Smee, J. [Niagara Technical Consultants, St. Catharines, Ontario (Canada)

2010-07-01

Gadolinium nitrate is added to the moderator of CANDU units to maintain the reactor in a guaranteed shutdown state (GSS). In April 2008, after being in stable GSS for over 30 hours, one of Ontario Power Generation's Pickering-B units showed a gradual depletion of the dissolved gadolinium, despite purification being isolated. Further additions of gadolinium stabilized the moderator gadolinium concentration, however, since the root cause of the depletion was not immediately identified, the unit was placed in the drained shutdown state, per established procedures. The cumulative gadolinium depletion amounted to about 3200 grams, the equivalent of about 12 ppm. Analysis showed the presence of oxalate in the moderator water. It is well-known that gadolinium forms a very insoluble oxalate (log K{sub sp} = -29.1). Although sub-micron filtration of water samples did not show the presence of gadolinium particulate, the measured levels of oxalate, 1.2 to 2 ppm, were sufficient to react with 1.4 to 2.4 ppm of gadolinium. The source of oxalate was traced to radiolysis of dissolved CO{sub 2} species. This unit had been experiencing chronic low-level ingress of CO{sub 2} from the Annulus Gas System. Free oxalate ion is normally susceptible to radiolytic breakdown back to CO{sub 2}, but Gd{sup 3+} provides a stable sink for radiogenic oxalate, 2 Gd{sup 3+} + 3 C{sub 2}O{sub 4}{sup 2-} → Gd{sub 2}(C{sub 2}O{sub 4}){sub 3}. Subsequent testing confirmed that gadolinium oxalate is quite stable with respect to gamma irradiation. Inspections showed well-crystallized gadolinium oxalate deposited on moderator system surfaces. Estimates indicated that about 1200 grams of gadolinium could have deposited on in-core surfaces, including the outside of the calandria tubes. That amount of negative reactivity was a concern, since it would prevent re-start of the unit. OPG, with support from AECL-Chalk River and Kinectrics, embarked on a two-pronged chemistry recovery program aimed at 1

A new CANDU-600 containment structure

International Nuclear Information System (INIS)

Serban, V.; Bobei, M.; Gheorghiu, M.; Popescu, M.; Stanciu, M.; Dinica, D.; Alexandru, C.

1994-01-01

This paper is presenting a structure made of reinforced concrete with rectangular cross-section, box-divided, prefabricated and modulled on a bay 6.5 m wide and 4.5 m high, and provided with a steel liner. The building has an overall basement in which the steel liner is embedded and which is supporting the building walls. The inner structure is common to the containment as well and it is carried out for each room (generally 6.5 m by 6.5 m) having intermediar floors at the necessary elevations. The containment dimensions, on horizontal plane are 6 x 6.5 m by 5 x 6.5 m and the total height of the side walls is 30.5 m. The containment is closed in A-C direction by a prefabricated semi-cylinder which is supported by the side walls and 5 intermediate arches. The fuel transfer deck structure is common to the inner structure and the containment structure. The Calandria vault is a separate individual structure located above E1. 100. For CANDU-600 main equipment the same arrangement was maintained, some unsignificant modifications being made, for example, the access areas located in the four corners of the building as well as the location of some auxiliary systems. The paper is also including a set of 1:200 scale drawings, comments on the construction manner and the results of the building structural analysis. The suggested solution is evidencing economical benefits facilities in the operation and construction of the plant and it is specially recommended for areas with high seismic events. (author)

Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

International Nuclear Information System (INIS)

Kim, Yonghee

2015-01-01

LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

Microstructure and crystallographic texture evolution during TIG welding of zircaloy-2 material

International Nuclear Information System (INIS)

Jha, S.K.; Singh, R.P.; Singh, V.K.; Ramanathan, R.; Samjdar, I.; Srivastava, D.; Tewari, R.; Dey, G.K.

2005-01-01

Zirconium and its alloys are extensively used as structural materials in nuclear reactors, because of better neutron economy, good corrosion resistance in water and good mechanical properties at operating temperature. Zircaloy-2 and zircaloy-4 are widely used in both pressurized water reactors (PWR) and boiling water reactors (BWR) as fuel cladding materials and as calandria tube and pressure tube materials in pressurized heavy water reactors (PHWR). The satisfactory performance and the life of the reactor components depend mainly upon their mechanical properties, corrosion properties and dimensional stability in the reactor condition, which are strong function of metallurgical parameters such as microstructure and texture. Therefore, for best performance of the reactor components these parameters are optimized during their fabrication. The microstructure and texture of the zircaloy-2 components are expected to get modified during the welding of the components. In this study the evolution of the microstructure and texture has been investigated as a function of the welding parameters. Heat input was varied the current and welding time. A variety of analytical techniques have been applied for the study on microstructure and texture of the welds. Optical microscopy and electron microscopy were used to evaluate the detailed microstructure. X-ray diffraction (XRD) was used investigate the crystallographic textures among the base metal, heat affected zone and fusion zone. Particular attention was focused on the determination of microtexture in weld by using electron backscatter diffraction (EBSD) technique. After that, an effort was put to compare the results of X-ray macro-texture and EBS-microtexture. (author)

Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2010-07-01

SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

Stress-induced reorientation of hydride precipitates in Zr-2.5Nb-0.5Cu garter springs under complex loading

International Nuclear Information System (INIS)

De, P.K.; John, J.T.; Raman, V.V.; Banerjee, S.

1991-01-01

Zr-2.5Nb-0.5Cu garter springs which are placed between coolant and calandria tubes in PHWRs experience complex loading due to simultaneous application of tension, compression and torus bending moment due to coolant tubes. The gradual pick up of hydrogen by the garter springs during service is likely to have hydride platelets reoriented under the applied stresses. In the present paper, the magnitudes and the directions of the principal stresses under the complex loading condition obtained have been calculated and the extent of hydride reorientation predicted. Simulation experiments consisting of simulated loading of hydrogen (upto 400 ppm) precharged springs at the service temperature (300degC) and also in-situ hydrogen charging of the springs under simulated loading conditions have been carried out. In addition, hydrogen precharged springs have been subjected to temperature cycling between 50 and 300degC under complex loading conditions, to evaluate the influence of temperature variation on hydride reorientation. Metallographic examination of the hydride platelets in the above springs has shown an excellent agreement with the analytical prediction. Torus bending moment values appear to play a significant role in reorienting the hydride platelets. It has been observed that under normal torus bending moment corresponding to 90 mm dia coolant tubes hydrogen platelets close to the outer rim of the spiral get reoriented in the radial direction. However, on application a torus bending moment corresponding to 30 mm dia tubes, hydride platelets get reoriented along the radial direction, irrespective of the magnitude of tensile and compression loading. (author). 9 refs., 15 figs., 1 appendix

The CIRENE program: experience gained in development of technology

International Nuclear Information System (INIS)

Villani, A.

1982-01-01

The construction of the Cirene 40 MW prototype electrical power plant at Latina is the main objective of the Cirene development program. A plant contract was given to NIRA at the end of 1976, and the plant completion is foreseen by the end of 1984. The Cirene is a heavy water-moderated, natural uranium-fueled pressure-tube reactor using boiling light water as the primary coolant. The design of Cirene is outlined. The numbers of pressure tubes, calandria tubes, liquid rod tubes and regulating rod tubes are 60, 60, 10 and 4, respectively, all made of Zircaloy-2. The equivalent diameter, length and lattice pitch of the core are 236 cm, 400 cm and 27 cm, respectively. The heavy water tank has the central zone of I.D. 369 cm and the dump annulus of I.D. 520 cm, made of AISI 304 L. These key parameters are shown. The means of regulating radioactivity during the normal operation of the reactor are: the two-phase rods, the level of the moderator and the concentration of boron in moderator. The function of two-phase rod regulation system is to introduce a variable density two-phase mixture consisting of boric acid solution and nonabsorbent gas in the core. The R and D was conducted to assess the functional and dynamic features of this control system. The plant is equipped with two independent fast-acting scram systems, discharging of moderator and liquid-rod system. The pressure-tube rupture test was carried out. (Nakai, Y.)

DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

Directory of Open Access Journals (Sweden)

Constantin POPESCU

2017-05-01

Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

Development and validation of a model for high pressure liquid poison injection for CANDU-6 shutdown system no.2

International Nuclear Information System (INIS)

Rhee, B.-W.; Jeong, C.J.; Choi, J.H.; Yoo, S.-Y.

2002-01-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the calandria tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, an AEA Technology CFD code, to simulate the formation and growth of the poison jet curtain inside the moderator tank. For validation, the current model is validated against a poison injection experiment performed at BARC, India and another poison jet experiment for Generic CANDU-6 performed at AECL, Canada. In conclusion this set of models is considered to predict the experimental results in a physically reasonable and consistent manner. (author)

Radioactive gaseous waste management activities at CWMF

International Nuclear Information System (INIS)

Sumangala, R.K; Cheralathan, M.; Hariharan, P.T.; Chitra, S.; Paul, B.

2015-01-01

HEPA and iodine filter banks are used as an important engineering safeguard to prevent the release of airborne activity to the environment during normal and accident conditions in all nuclear installations. CWMF is responsible for the periodical testing and certification these filter banks as per the technical specification of the nuclear facilities at Kalpakkam site. An efficiency of >99.9% is ensured for both the HEPA as well as iodine filter banks. The larger radioactive particulates are trapped in the micro glass fibre filter paper medium by the mechanism of interception and inertial impaction whereas particulates of submicron size are caught by diffusion. The major activity removed in particulate form is 137 Cs and 90 Sr. The elemental iodine is removed by physico-chemical adsorption on high surface area activated charcoal and organic compounds of iodine are removed by isotopic exchange with KI/KOH impregnated activated charcoal or silver impregnated silica gel. Silver impregnated molecular sieves 13-X and AR-1 were developed for the removal of iodine from reprocessing atmosphere. Studies on pressure swing adsorption technique have been carried out for isolating Argon from air. Using Molecular sieve 5A (45psi-50psi) and Carbon molecular sieves (100psi to 120psi) based PSA systems in series an enrichment of 30% Ar is possible. Pressurized Heavy Water Reactors with calandria vault coolant as air produces 41 Ar due to neutron activation of 40 Ar present in air which main contributor to the air is borne activity in MAPS, RAPS and Dhruva reactors. The isolated argon can be stored for decay and activity release can be minimized as per ALARA principle. (author)

2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

Energy Technology Data Exchange (ETDEWEB)

Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

2016-12-01

Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

International Nuclear Information System (INIS)

Inozemtsev, V.; Ganguly, C.

2005-01-01

Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

Experience in the application of the IAEA QA code and guides to the manufacture of nuclear reactor components

International Nuclear Information System (INIS)

Dutta, N.G.; Mankame, M.A.; Kulkarni, P.G.; Vijayaraghavan, R.; Balaramamoorthy, K.

1985-01-01

India has made considerable progress in the indigenous manufacture of 'Quality' nuclear reactor components. All activities associated with the development of atomic energy from mining of strategic minerals to the design, construction, and operation of nuclear power plants including supporting research and development efforts are mainly carried out by the Department of Atomic Energy (DAE). Through the sustained efforts of DAE, the major industries, both in public and private sectors supplying nuclear components have now adopted the practice of systematic quality assurance (QA). The stringent QA steps are mandatory for achieving the desired quality in the manufactured nuclear components. Control blades for BWRs are now indigenously manufactured by the Atomic Fuels Division (AFD) of Bhabha Atomic Research Centre (BARC), a constituent unit of DAE. For the Project Dhruva, a 100 MW(th) nuclear reactor, constructed at BARC, Trombay, Bombay, an independent cell was formed to carry out quality audit on the manufactured components. The components were designed, fabricated, inspected and tested to the desired quality level. The QA activities were enforced from the procurement of raw materials to the audit of the completed component for monitoring the manufacturer's continued compliance with the design. The major components of Dhruva, viz. calandria, end-shield, coolant channels, heat exchangers, etc., were covered under these quality audit activities. The paper highlights the QA programme implemented in the manufacture of control blades for BWRs, illustrated with a typical example, the end-shield for Dhruva. The authors consider that the recommendations and guidelines provided in the documents 50-SG-QA3, 50-SG-QA8, 50-SG-QA10, etc., were useful in providing a formal and systematic framework, under which various quality assurance functions have been carried out

Modelling disassembled fuel bundles using CATHENA MOD-3.5a under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Lei, Q M; Sanderson, D B; Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

CATHENA MOD-3.5a is a multipurpose thermalhydraulic computer code developed primarily to analyse postulated loss-of-coolant scenarios for CANDU nuclear reactors. The code contains a generalized heat transfer package that enables it to model the behaviour of a fuel channel in great detail. Throughout the development of the CATHENA code, considerable effort has been devoted to evaluating, validating and documenting its overall capability as a design and safety assessment tool. Specific attention has focused on its ability to predict fuel channel behaviour under postulated accident conditions. This paper describes an investigation of CATHENA`s ability to predict the thermal-chemical responses of a fuel channel in which the 37-element bundles were assumed to disassemble and rearrange into a closed-packed stack of elements at the bottom of the pressure tube. A representative disassembled bundle geometry was modelled during a simulated loss-of-coolant accident scenario using CATHENA MOD-3.5a/Rev 0, with superheated steam being the only coolant available. Thermal conduction in the radial and circumferential directions was calculated for individual fuel elements, the pressure tube, and the calandria tube. Radiation view factors for the intact and disassembled bundle geometries were calculated using a CATHENA utility program. Inter-element metal-to-metal contact was accounted for using the CATHENA solid-solid contact model. An offset pressure-tube configuration, representing a partially sagged pressure tube, and the effect of steam starvation on the exothermic zirconium-steam reaction, were included in the CATHENA model. The CATHENA-predicted results show a dramatic suppression of heat generation from the zirconium-steam reaction when bundle disassembly is initiated. The predicted results show a smaller temperature increase in the fuel sheaths and the pressure tube for the disassembled bundle geometry, compared to the temperature excursion for the intact bundle. (author

Development of transducers for integrated garter spring repositioning system

Energy Technology Data Exchange (ETDEWEB)

Sharma, B S.V.G.; Shyam, T V; Shrivastava, A K; Rupani, B B; Sinha, R K [Bhabha Atomic Research Centre, Bombay (India). Reactor Engineering Div.

1994-12-31

In order to reposition the dislocated garter springs in active channels of 235 MW Pressurised Heavy Water Reactors (PHWRs), a tool named as Integrated Garter Spring Repositioning System (INGRES) has been developed. The tool consists of transducers to detect the concentricity between the Pressure Tube (P/T) and Calandria Tube (C/T) and also to detect garter springs in the channel besides different modules for correcting the eccentricity between P/T and C/T and garter spring repositioning. The transducers used in the system namely Concentricity Detection Probe (CDP) and Garter Spring Detection Probe (GSDP) are based on the eddy current techniques. The CDP makes use of four eddy current bobbin probes separated 90 degrees apart in cross sectional plane of channel assembly. The transducer gives output signal in proportional to the air gap between P/T and C/T in two axes (X and Y) which are designed for the purpose. The output of the unit is obtained on the Cathode Ray Oscilloscope (CRO) screen in the form of illuminated dot. The dot position on the CRO screen gives the information about mismatch in concentricity between P/T and C/T of the channel. The GSDP meant for detecting garter springs in PHWR channel uses two sets of primary and secondary coils connected in differential mode. The output signals from the transducers are processed through a signal processing unit devised for the purpose to obtain output from it as a horizontal beam on the CRO screen. The garter spring presence in the channel is indicated by a change in the voltage level of beam and also by audio-visual indication in the form of buzzer and LED illumination on the processing unit. This paper gives general design and development aspects of the CDP and GSDP transducers of the INGRES tool. (author). 3 figs.

Zirconium - an imported mineral commodity

International Nuclear Information System (INIS)

1983-10-01

This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

Effective utilization of maintenance staff in design and implementation of major project work

International Nuclear Information System (INIS)

Wyman, D.; Dingle, J.; Brown, R.

1995-01-01

The reorganization of Pickering Nuclear Division some 2 years ago resulted in the formation of the Projects and Modifications department. This department takes an integrated approach to manage all aspects of large projects at Pickering. The integration of Design, Drafting, Procurement, Construction and Operations functions into project teams represents a fundamental change to project management at Pickering. The development of integrated teams has great potential for reducing both the time and cost associated with project implementation, while at the same time improving the quality, and maintainability of the commissioned in service project. The Pickering Rehab organization 1989-1993, established to perform the rehab / retube of Units 3 and 4 had proven that a team environment will produce effective results. The outcome was astounding, critical categories such as Safety, Quality of Work, and Timeliness, had proven the team's effectiveness. The integration of operations maintenance staff into the project work activities is still evolving, and has probably required the most adaptation to change for both the former Construction and Operations organizations. Maximizing the utilization of the maintenance staff in the design and implementation of major project work will prove to be a key to a long term operating success of these projects. This paper will focus in on the effective usage of Maintenance staff in the design and implementation phases of major project work at Pickering, and on the benefits realized using this approach. It will be divided into 5 sections as indicated. 1. Past Project Shortfalls. 2. Benefits of the inclusion of Maintenance staff in the Calandria Vault Rehab Project. 3. Maintenance involvement in the Pickering 'A' Shutdown System Enhancement (SDSE) Project. 4. Challenges resulting from the inclusion of Maintenance staff project teams. 5. Summary. (author)

Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.N.; Mohan, L.R.

2002-01-01

Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

Ignitability of hydrogen/oxygen/diluent mixtures in the presence of hot surfaces

International Nuclear Information System (INIS)

Kumar, R.K.; Koroll, G.W.

1995-01-01

In the licensing process for CANDU nuclear power stations it is necessary to demonstrate tolerance to a wide range of low-probability accidents. These include loss of moderator accidents that may lead to the formation of flammable mixtures of deuterium, oxygen, helium, and steam in the reactor calandria vessel. Uncovered adjuster or control rods are considered as possible sources of ignition when a flammable mixture is present. A knowledge of the minimum hot-surface temperature required for ignition is important in assessing the reactor safety. These hot surface temperatures were measured using electrically heated adjuster rod simulators in a large spherical vessel (2.3-m internal diameter). Whereas the effects of geometry on ignition temperature were studied in the large-scale apparatus, some of the effects, such as those produced by a strong radiation field, were studied using a small-scale apparatus. Investigations carried our over a range of hydrogen and diluent concentrations indicated that, although the ignition temperatures were fairly insensitive to the hydrogen concentration, they were strongly affected by the presence of steam The addition of 30% steam to a dry combustible mixture increased the minimum surface temperature required for ignition by approximates 100 degrees C of the diluents investigated, steam had the most effect on ignition. The effect of initial temperature of the mixture on the ignition temperature was small, whereas the effect of initial pressure was significant. The effect of substituting deuterium for hydrogen on ignition temperature was small. The effect of a high-intensity gamma-radiation field on the minimum hot-surface temperature required for ignition was investigated using a 2-dm 3 ignition vessel placed in a linear accelerator. Radiation had no measurable effect on ignition temperature

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

Gavrilas, M.; Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1994-01-01

The GOTHIC code was employed to assess the effectiveness of several original heat rejection features that make it possible to cool large rating containments. The code was first verified and modified for specific containment cooling applications; optimal mesh sizes, computational time steps, and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. GOTHIC was then used to obtain performance predictions for two containment concepts: a 1200 MW e new pressure tube light water reactor, and a 1300 MW e pressurized water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features have been predicted. For the 1200 MW e pressure tube light water reactor, the evaluated pressure-limiting features are: a large water pool connected to the calandria, large containment free volume and an air-convection annulus. For the 1300 MW e pressurized water reactor, an external moat, an internal water pool, and an air-convection annulus were evaluated. The performance of the proposed containment configurations is dependent on the extent of thermal stratification inside the containment. The best-performance configurations/worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light water reactor, and less than 0.45 MPa for the 1300 MW e pressurized water reactor. The low peak pressure predicted for the 1200 MW e pressure tube light water reactor can be in part attributed to its relatively large free volume, while the relatively high peak pressure predicted for the 1300 MW e pressurized water reactor can be attributed to its relatively small free volume (i.e., the size used was that of a pressurized water reactor containment designed with active heat removal features). (author)

Numerical modelling of the impedance plane for simultaneous determining, by eddy currents, the oxide thickness and conductivity in Zircaloy

International Nuclear Information System (INIS)

Lois, Alejandro E.

2001-01-01

During service at high temperature or in aggressive media, metallic structures and components may suffer different types of changes or degradation. As an example, phase transformations may occur, second phases may precipitate, and in consequence the mechanical and chemical properties of the material may change. Their behavior will therefore differ from that considered for the design of the component. The knowledge of the amount of a precipitated second phase in a component should be an important tool in the hands of the maintenance engineer. And it would be very important to obtain this knowledge nondestructively and reliably. The objective of this project is to evaluate by eddy currents the amount of the hydrogen incorporated during service in structural zirconium base materials. For this purpose, a series of Zircaloy-4 specimens with oxide layers of different thickness and different concentration of hydrogen obtained by controlled autoclave treatments were used. These specimens were tested with eddy current equipment. The information produced by an eddy current test is the superposition of many variables, e.g.: thickness of oxide layers, conductance, thickness of specimen, etc. In order to sort out this information, an analytical model of the impedance plane was programmed in a PC, with which this information was processed, permitting, in this way, to evaluate the conductivity of materials, taking into account the effect of oxide layers thickness. A linear relationship between the conductivity and the hydrogen content in the range of hydrogen concentrations of technological interest was observed. Therefore, the calculated electrical conductivity may be transformed to the amount of hydrogen content, using a suitable calibration curve. This process will allow for the nondestructive assessment of the amount of hydrogen in reactor components, such as pressure and calandria tubes, a knowledge which will enable the experts to predict the degree of fragility of those

A rationale for the observed non-linearity in pressure tube creep sag with time in service

International Nuclear Information System (INIS)

Sedran, P.J.

2013-01-01

In 2012, a paper was presented at the CNS SGC Conference which included an explanation for measured non-linear trends in Pressure Tube (PT) creep sag. The section of the 2012 paper covering this topic was revised and is presented as the main subject of this paper. The practical applications for the prediction of long-term Fuel Channel (FC) creep sag include the analysis of Calandria Tube - Liquid Injection Nozzle (CT-LIN) contact, and fuel passage and PT replacement assessments. The current practice for predicting FC creep sag in life cycle management applications is to use a linear model for creep sag versus time in service. However, PT sag measurements from the Point Lepreau Generating Station (PLGS) and Gentilly-2 (G-2) have displayed a non-linear trend with a creep sag rate that is decreasing with time in service. As an example, for PT F06 in PLGS, a 60% reduction in the nominal creep sag rate was observed for measurements taken 18 years apart. Subsequently, it was found that a 56% reduction in the creep sag rate for F06 over 18 years could be attributed to a fundamental geometric property of the PT creep sag profile. In addition, a further 1.6% decrease in the creep sag rate of the CT over the same period could be attributed to bending stress reductions due to the deformation of the CT. The resultant reduction in the PT creep sag rate for F06 was predicted to be 57.6%, closely matching the observed PT creep sag rate reduction of 60%. Therefore, this paper provides a rationale to explain the observed non-linear trends in PT creep sag, the use of which could benefit stations engaging in asset management as a means of FC life extension. This paper presents a summary of the worked performed to correlate the observed reductions in PT creep sag rate to the geometrical properties of the PT creep sag profile and the predicted bending stress reductions in the CT. (author)

Probabilistic evaluations for CANTUP computer code analysis improvement

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2004-01-01

Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

International Nuclear Information System (INIS)

Rupani, B.B.; Sinha, S.K.; Sinha, R.K.

2002-01-01

Zirconium alloy coolant channels are central to the design of Indian Pressurised Heavy Water Reactors (PHWRs) and form the individual pressure boundaries. These coolant channels consist of horizontal pressure tubes made of zirconium alloys, which are separated from cold calandria tubes using garter spring spacers. High temperature heavy water coolant flows through the pressure tube which supports the fuel bundles. A typical coolant channel in a PHWR is shown. These pressure tubes are subjected to several life limiting degradation mechanisms like creep and growth, hydrogen pick-up, reduction in fracture toughness and delayed hydride cracking phenomena because of their operation under high temperature, high stress and high fast neutron flux environment. Considering the early onset of these degradation mechanisms in Zircaloy-2 pressure tubes used in the early generation of Indian PHWRs, the life management of these coolant channels becomes a challenging task, involving multidisciplinary R and D efforts in areas like analytical modelling of degradation mechanisms, evolution of methodologies for assessment of fitness for service and, tools and techniques for remote on line monitoring of integrity, maintenance and replacement. The degradation mechanisms have been modelled and incorporated into specially developed computer codes, such as SCAPCA for irradiation induced creep and growth deformation modelling, HYCON for hydrogen pick-up modelling, BLIST for hydrogen diffusion, blister nucleation and growth modelling and CEAL for assessment of leak before break behaviour. These codes have been validated with respect to the results of in-service inspection and post irradiation examination. Development of analytical models actually paved the way for the evolution of more refined methodologies for assessing the safe residual life of coolant channel. Information gathered from various experiments simulating the degradation mechanisms, results of post-irradiation examination of the

Integrated planning for a fuel industry with emphasis on minimum size to fabricate own fuel

International Nuclear Information System (INIS)

Kondal Rao, N.; Katiyar, H.C.; Rajendran, R.; Sinha, K.K.; Swaminathan, N.; Subramanyam, R.B.; Pande, B.P.; Krishnan, T.S.; Agarwala, G.C.; Chandramouli, V.A.

1977-01-01

The Indian nuclear energy programme is based on the utilization of indigenous resources for the economic generation of power, developing its own know-how. In order to gain time, the first nuclear power station at Tarapur is a turn-key job based on enriched uranium fuel. Taking into consideration the established resources of uranium and thorium in the country, a strategy for nuclear power programme has been drawn up. The first phase is based on natural uranium fuel, the second phase on the recycle of plutonium and conversion of thorium and the third phase is the breeder system based on utilization of U 233 and conversion of thorium. This programme is specially significant for India in view of its vast resources of thorium. After the experience and confidence gained with the manufacture of metallic uranium fuel for the research reactors and about 40 tonnes of fuel for the initial loading of the Rajasthan Reactor, the fuel manufacturing programme within the country has been implemented to meet the entire initial and reload fuel requirements. The plant capacities are small compared to similar activities in developed countries. Further, by planning for an integrated fuel and component manufacturing complex, any draw-back in smaller scale of some of the operations is off-set. At the Nuclear Fuel Complex, set up on the above principles, production plants are in operation for the manufacture of reload fuel for the 400 MW Tarapur station, natural uranium oxide fuel, various zircaloy components such as fuel sheaths, pressure tubes, calandria tubes, channels and various other zircaloy components. Provisions have been made to expand the production facilities as the demand for reload fuel grows. With the facilities provided, the production programme can be diversified to take up the production of fast breeder reactor components of stainless steel and also the blanket thorium elements. The unitary control of all aspects of the manufacture and quality control of different types

SLARette Mark 2 System

International Nuclear Information System (INIS)

Wills, J.L.

1991-01-01

The SLAR (Spacer Location and Repositioning) program was initiated to develop the technology to locate and reposition the fuel channel spacers that separate the pressure tube from the calandria tube. The requirements for SLAR are for the system to be operable on channels with up to 100,000 EFPH (effective full power hours) and to be used on a continuous basis in an automated mode. The SLAR system was therefore developed based on existing fuelling machine technology. The CANDU 6 SLAR Delivery Machine is shown, it is sightly larger than a CANDU 6 Fuelling Machine. The SLAR delivery machine contains a mechanical ram which removes the channel closure and shield plugs and a telescopic hydraulic ram which deploys the SLAR Tool into the fuel channel, these two rams are indexed by means of a turret which is attached to a conventional fuelling machine magazine and snout assembly. A large drum is located beneath the magazine to take up the umbilical cable which supplies the SLAR Tool. SLAR requires the removal of a fuelling machine from the fuelling machine bridge which is replaced by the SLAR Delivery Machine. The SLAR Delivery Machine works in conjunction with the other fuelling machine to defuel the channel and in effect becomes part of the fuel handling system with the operation run from the fuel handling control console. The SLAR operating system is shown. It is expected that the SLARing of a complete reactor in this manner will take 38 months (depending on the amount of spacer movement needed). The SLARette system evolved from SLAR in response to a need by some utilities to avoid the long outage associated with SLAR and achieve the same results over several annual (short) outages. The basic requirements for SLARette were therefore still to be operable on channels with up to 100,000 EFPK to make maximum use of developed SLAR technology and to be quick and easy to install and remove. The SLARette system as described below meets these requirements. SLARette utilizes

La lumière et le temps sur la scène baroque : Poetique & Pratique

Directory of Open Access Journals (Sweden)

Françoise Siguret

2016-06-01

Full Text Available Time: Aristotle, in the Poetics, recommends the playwright to confine his tragedy within «two revolutions of the sun»; the concept refers to the light perception, to the fact that greek drama is acted in the open air. The messengers and the chorus represented on the stage, in the present time, what happened outside of it. In the age of the French classical theatre, the chronological sequence of the action had to conform to the laws of the reason: the so-called rule of the twenty-four hours became an indisputable rule of the action. A time exactly measured, substituting the time of the light, cyclical and mythical. In Italy, pastorals, mythological melodramas and all that belonged to the court entertainments (ballets, operas, tournaments conformed to a cyclical time in which the four seasons constituted the scenery, linking life to the four liturgical seasons and to the four parts of the day, from noon to midnight (cfr. Endymion and the Ballet de la Nuit. Light: Need to light up the indoor playhouse for practical and moral issues. Italian craftsmen implement the technical tools; a certain difference between primary light (intended to light up the stage and the auditorium and the lumi (the supplementary lighting related to a specific performance. Buontalenti’s lighting devices (sun, moon, rainbows, divine and princely splendour will enchant the spectators. France will discover these stagecraft effects with the Calandria (1548, without subsequent developments. Afterwards, Corneille will be fascinated by the ‘baroque’ charm (Médée, 1639 and Andromède, 1650. In the second half of the XVIIth century, while machinery invades opera and tragedy in music, Racine refuses anything intended to deceive the eye, though creates a lighting that may be «listened» (Britannicus. The Allegories (the «other discourse» convey meanings on the baroque stage through the perpetual slow motion of the gods and Time, till the final glory of the Prince: Cosimo

Point Lepreau refurbishment - update 5

International Nuclear Information System (INIS)

White, R.M.; Eagles, E.R.; Hickman, C.N.; Baker, R.; Thompson, P.D.; Howieson, J.Q.; Ichiyen, N.

2005-01-01

NB Power Nuclear is planning to conduct a 18-month maintenance outage of the Point Lepreau Generating Station (PLGS) beginning in April 2008. The major activity would be the replacement of all 380 Fuel Channel and Calandria Tube assemblies and connecting feeder pipes. This activity is referred to as Retube. NB Power Nuclear would also take advantage of this outage to conduct a number of repairs, replacements, inspection and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc.). These collective activities are referred to as Refurbishment. This would allow the station to operate for an additional 25 to 30 years. The scope or the project was determined [mm the outcome of a two year study involving a detailed condition assessment of the station which examined Issues relating to ageing and obsolescence, along with a detailed review or Safety and Licensing issues associated with extended operation. The Refurbishment outage would be preceded by a detailed Engineering Project Phase that would: Finalize details of the Retube process including modeling, tooling development, site facilities and training of personnel. Perform necessary engineering activities related to design modifications. Construct the new waste storage structures to house Retube Waste and other additional waste storage structures for the extended life of the station. Setup necessary temporary construction facilities (offices, storage areas, change moms, decontamination an maintenance areas) to support Retube. Procure equipment and components. Perform detailed outage planning. Initiate development of detailed commissioning as well as lay-up, monitoring and return to service procedures. At the present time, the NB Power Nuclear Board of Directors and the New Brunswick Provincial government are reviewing a proposal for a lease arrangement from Bruce Power

Technologies in support of CANDU development

International Nuclear Information System (INIS)

Turner, C.; Tapping, B.

2005-01-01

Atomic Energy of Canada, Ltd. (AECL) has significant research and development (R and D) programs designed to meet the needs of both existing CANDU reactors and new and evolving CANDU plant designs. These R and D programs cover a wide range of technology, from chemistry and materials support through to inspection and life management tools. Emphasis is placed on effective technology development programs for fuel channels, feeders and steam generators to ensure their operation through design life, and beyond. This paper specifically addresses how the R and D has been applied in the production of longer-lived pressure tubes for the most recent CANDU 6 reactors, and how this technology forms the basis for the pressure tubes of the Advanced CANDU Reactor (ACR). Similarly, AECL has developed solutions for other critical components such as calandria tubes, feeder pipe and steam generators. The paper also discusses how the R and D knowledge has been integrated into aging management databases and health monitoring tools. Since 1997, AECL has been working with CANDU utilities on comprehensive and integrated CANDU Plant Life Management (PLiM) programs for successful and reliable plant operation through design life and beyond. AECL has developed and implemented an advanced chemistry monitoring and diagnostic system, called ChemAND which allows on-line access by the operators to current and past chemistry conditions enabling appropriate responses and facilitating planning of shutdown maintenance actions. An equivalent tool for monitoring, trending and diagnosing thermal and mechanical data has also been developed; this tool is called ThermAND. AECL is developing the Maintenance Information, Monitoring, and Control (MIMC) system, which provide information to the user for condition-based decision-making in maintenance. To enable more effective inspections, surveillance and data collection, AECL has developed unique one-off tooling to carry out unanticipated inspection and repair

Possible refurbishment of Point Lepreau

International Nuclear Information System (INIS)

White, R.M.; Groom, S.H.; Thompson, P.D.; Barclay, J.M.; Allen, P.J.

2001-01-01

In February 2000, the NB Power Board of Directors approved Phase one of a project to produce a business case including a detailed scope and estimate associated with the possible refurbishment of the Point Lepreau Generating Station (PLGS). The Preliminary plan for refurbishment projects an 18-month outage starting as early as the spring of 2006. If the station were to be refurbished, then it would be run for another 25 to 30 years. The decision on whether or not to refurbish PLGS has not been made and is not expected until the summer of 2002. The results of the first phase of the project will be used to prepare a detailed business case that will be presented to the NB Power board of directors in January of 2002. At that time a decision will be made as to whether to refurbish the unit, or obtain other means of replacing the energy produced by PLGS. The station currently produces about a third of the power generated within the province. If the business case is approved, all-380 Pressure Tubes and Calandria Tubes, along with their related End Fittings and Feeders would be replaced. This material would be stored in new storage vaults to be constructed at the existing on-site Waste Management Facility. Replacement of other station components will be performed as required, as determined from the results of a comprehensive Plant Condition Assessment. The condition assessments build on work done under the Plant Life Management Program. Point Lepreau Generating Station has operated well since start of commercial operation in early 1983. With a lifetime capacity factor of about 84% (up to the end of 2000), it has proven to be an economic and environmentally sound electricity provider. The station has also had a significant positive economic impact in Southern New Brunswick, employing over 600 people. However the Pressure Tubes and Feeders are nearing the point in time in which they will exceed their fitness for service criteria. Although tubes can be replaced on an

Possible refurbishment of Point Lepreau

International Nuclear Information System (INIS)

White, R.M.; Groom, S.H.; Thompson, P.D.; Barclay, J.M.; Allen, P.J.

2001-01-01

In February 2000, the NB Power Board of Directors approved Phase one of a project to produce a business case including a detailed scope and estimate associated with the possible refurbishment of the Point Lepreau Generating Station (PLGS). The Preliminary plan for refurbishment projects an 18-month outage starting as early as the spring of 2006. If the station were to be refurbished, then it would be run for another 25 to 30 years. The decision on whether or not to refurbish PLGS has not been made and is not expected until the summer of 2002. The results of the first phase of the project will be used to prepare a detailed business case that will be presented to the NB Power board of directors in January of 2002. At that time a decision will be made as to whether to refurbish the unit, or obtain other means of replacing the energy produced by PLGS. The station currently produces about a third of the power generated within the province. If the business case is approved, all-380 Pressure Tubes and Calandria Tubes, along with their related End Fittings and Feeders would be replaced. This material would be stored in new storage vaults to be constructed at the existing on-site Waste Management Facility. Replacement of other station components will be performed as required, as determined from the results of a comprehensive Plant Condition Assessment. The condition assessments build on work done under the Plant Life Management Program. Point Lepreau Generating Station has operated well since start of commercial operation in early 1983. With a lifetime capacity factor of about 84% (up to the end of 2000), it has proven to be an economic and environmentally sound electricity provider. The station has also had a significant positive economic impact in Southern New Brunswick, employing over 600 people. However the Pressure Tubes and Feeders are nearing the point in time in which they will exceed their fitness for service criteria. Although tubes can be replaced on an

Process improvements for enhanced productivity of PHWR garter springs

International Nuclear Information System (INIS)

Srinivasula Reddy, S.; Tonpe, Sunil; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Full text: In Pressurised Heavy Water Reactors (PHWR), Garter springs are used as spacers between the coolant tube and calandria tube. Garter springs are made from Zirconium alloy containing 2.5 % Niobium and 0.5% copper. The springs are basically manufactured by coiling a wire of cross section 1.7 mm x 1.0 mm, which is produced by series of drawing and swaging operations using hot extruded rods of 19 mm diameter. The manufacturing process also involves heat treatment and chemical cleaning operations at appropriate stages. It is required to ensure that the life of springs against parameters like hydrogen pickup, residual stresses and low stiffness is improved at the manufacturing stage itself by improving manufacturing process. The impact of above problems on spring life and process improvements is briefly discussed. The critical factor affecting the garter spring performance in PHWR Reactor is mainly hydrogen. The life limiting factors for garter springs are the problems arising out of high total hydrogen content, which depends on the hydrogen pickup during reactor operation. This phenomenon can happen during the reactor operation, as springs are prone to pick-up hydrogen in the reactor environment. Hence acceptable hydrogen content for the springs is specified as 25 ppm (max.). Garter spring is susceptible to hydrogen pick-up during various production processes, which make material brittle and difficult for fabrication process such as wire drawing and coiling. By studying and optimizing the process parameters of spring manufacturing, the hydrogen pick-up of springs is brought down from 70 ppm to a level of 20 ppm. Garter springs are provided with a hook at each end to enable its assembly to coolant tube in the reactor. The hook portion is very critical in maintaining the integrity of the spring. It is desirable to have the hook portion relieved of all residual stresses. For this purpose manufacturing process has been modified and solutionising was introduced as

R and D in support of CANDU plant life management

International Nuclear Information System (INIS)

Tapping, R.L.; Holt, R.A.

1999-01-01

One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

Fundamentos ecológicos y biogeográficos de la rareza de la avifauna madrileña: Una propuesta de modificación del catálogo regional de especies amenazadas

Directory of Open Access Journals (Sweden)

Carrascal, L. M.

2006-05-01

ático – ‘Sensible a la Alteración de su Hábitat’; Calandria – descatalogarla; Mirlo Acuático - ‘Sensible a la Alteración de su Hábitat’; Colirrojo Real - ‘Sensible a la Alteración de su Hábitat’ o ‘Vulnerable’; Papamoscas Gris - ‘De Interés Especial’; Alcaudón Real Meridional – descatalogarla; Picogordo – ‘De Interés Especial’.

Design of a PWR for long cycle and direct recycling of spent fuel

Energy Technology Data Exchange (ETDEWEB)

Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

2015-12-15

assembly, 58 or 54 MWd/kgU burnup of the fuel can be achieved, respectively. Comparing this fuel cycle strategy with that of the advanced pressurized water reactors such as AP-1000 and EPR, we find that almost the same burnup can be achieved with longer core cycle: 36 months versus 18 months for these reactors. Recycling these fuel bundles in CANDU-6 reactors requires minimizing the lattice pitch to 22 cm. In order to keep the coolant void reactivity of the ACR-700 slightly negative, the calandria tube should be increased from 7.8 cm to 8.2 cm to decrease the moderator to fuel volume ratio.

Lessons-learned from ongoing decommissioning project of Fugen NPS

International Nuclear Information System (INIS)

Tezuka, M.; Koda, Y.; Iguchi, Y.; Kato, Y.; Yanagihara, S.

2017-01-01

Advanced Thermal Reactor (ATR) Fugen is a 557 MWt, 165 MWe, heavy water moderated, light-water cooled, pressure-tube type reactor. In 2003, Fugen was shut down after ca 25 years operation, and started decommissioning activity from 2008. In the initial period of decommissioning, we have been dismantling from turbine systems because of their little contamination. In general, most difficult process of dismantling of nuclear power plant is the dismantlement of the reactor core because the radiation rate of the reactor core is very high, e.g., it is over 200 Sv/hr in the Fugen's case. Our plan of dismantlement of the core is from about 2022. The core area has some features that the structure is narrow and complicated by tube-cluster structure that contains 224 fuel channels with both the pressure and the calandria tubes coaxially in each channel. The radiation shielding area is laminated structure composed of up to 150 mm thickness of carbon steel. And the structure of the reactor, which is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum. In particular, the core area is planning to be dismantled under water by remote controlled machines in order to shield the radiation around the core and prevent airborne dust generated by the cutting considering the usage of Zr alloy which is likely to be oxidized. In consideration of above, the cutting methods were selected for dismantling the reactor core in order to shorten the dismantling term and reduce the secondary waste. The candidate cutting method options were decreased based on the results of the researches on achievement of the cutting methods domestically and internationally. Finally, the laser cutting method was selected for dismantling the core area and shielding area, and diamond wire saw was also selected for dismantling the shielding area applicable to concrete with metal liner, based on the results of some cutting tests. The laser cutting method has many advantages, e

Modelling flow and work hardening behaviour of cold worked Zr–2.5Nb pressure tube material in the temperature range of 30–600 oC

International Nuclear Information System (INIS)

Dureja, A.K.; Sinha, S.K.; Pawaskar, D.N.; Seshu, P.; Chakravartty, J.K.; Sinha, R.K.

2014-01-01

Under a postulated accident scenario of loss of cooling medium in an Indian Pressurised Heavy Water Reactor (IPHWR), temperature of the pressure tubes can rise and lead to large deformations. In order to investigate the modes of deformation of pressure tube – calandria tube assembly, material property data defining the flow behaviour over a temperature range from room temperature (RT) to 800 o C are needed. It is of practical importance to formulate mathematical equations to describe the stress–strain relationships of a material for a variety of reasons, such as the analysis of forming operations and the assessment of component's performance in service. A number of constitutive relations of empirical nature have been proposed and they have been found very suitable to describe the behaviour of a material. Although these relations are of empirical nature, various metallurgical factors appear to decide applicability of each of these relations. For example, grain size influences mainly the friction stress while the strain hardening is governed by dislocation density. In a recent work, tensile deformation behaviour of pressure tube material of IPHWR has been carried out over a range of temperature and strain rates (Dureja et al., 2011). It has been found that the strength parameters (yield and ultimate tensile strength) vary along the length of the tube with higher strength at the trailing end as compared to the leading end. This stems from cooling of the billet during the extrusion process which results in the variation of microstructure, texture and dislocation density from the leading to the trailing end. In addition, the variation in metallurgical parameters is also expected to influence the work hardening behaviour, which is known to control the plastic instability (related to uniform strain). In the present investigation, the tensile flow and work-hardening behaviour of a cold worked Zr–2.5Nb pressure tube material of IPHWRs has been studied over the

Refurbishment of Point Lepreau Generating Station

International Nuclear Information System (INIS)

Thompson, P.D.; Jaitly, R.; Ichiyen, N.; Petrilli, M.A.

2004-01-01

NB Power is planning to conduct an 18-month maintenance outage of the Point Lepreau Generating Station (PLGS) beginning in April 2008. The major activity would be the replacement of all 380 Fuel Channel and Calandria Tube Assemblies and the connecting feeder pipes. This activity is referred to as Retube. NB Power would also take advantage of this outage to conduct a number of repairs, replacements, inspections and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc). These collective activities are referred to as Refurbishment. This would allow the station to operate for an additional 25 to 30 years. The scope of the project was determined from the outcome of a two-year study involving a detailed condition assessment of the station that examined issues relating to ageing and obsolescence. The majority of the plant components were found to be capable of supporting extended operation without needing replacement or changes. In addition to the condition assessment, a detailed review of Safety and Licensing issues associated with extended operation was performed. This included a review of known regulatory and safety issues, comparison of the station against current codes and standards, and comparison of the station against safety related modifications made to more recent CANDU 6 units. Benefit cost analyses (BCA) were performed to assist the utility in determining which changes were appropriate to include in the project scope. As a Probabilistic Safety Assessment (PSA) for PLGS did not exist at the time, a risk baseline for the station had to be determined for use in the BCA. Extensive dialogue with the Canadian Nuclear Safety Commission staff was also undertaken during this phase. A comprehensive Licensing Framework was produced upon which the CNSC provided feedback to NB Power. This feedback was important in terms of

Clean energy for a new generation. Steam generator life cycle management and Bruce restart

International Nuclear Information System (INIS)

Newman, G.W.

2009-01-01

In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre