chapter 7 safety analysis and research · 78 aerb annual report - 2018 indicate full 2d structure...

73AERB Annual Report - 2018

AERB recognises the importance of Safety analysis & Research in support of its regulatory processes. In-house safety research and development (R&D) helps in obtaining deeper insights into the issues concerning nuclear and radiation safety to arrive at scientifically sound and risk informed regulatory decisions. Safety analysis and research activities are carried out by AERB as a part of its regulatory activities. Several important developmental studies were taken up by AERB and completed during this year. A brief overview of these activities is presented in the following sections.

7.1 THERMAL HYDRAULICS SAFETY STUDIES

7.1.1 Flow Reduction Studies in FBTR Special Subassembly (ISZ100)

A special subassembly (ISZ100) is used in FBTR to irradiate the ternary metal fuel (U-23wt%Pu-6wt%Zr) pins for assessing its behaviour under irradiation. As part of regulatory review, an independent verification study was carried out to assess the effects of partial flow blockage in ISZ100 to determine the maximum allowable flow reduction such that the fuel, clad and sodium temperatures do not exceed permissible limits. A 3-D CFD analysis of four cases, i.e., normal operating conditions (100% flow) and 40%, 60% and 74% flow blockage. It was found that the maximum allowable flow reduction is governed by clad temperature limit. The results from this analysis compared favourably with those provided by the utility. Temperature contour at the top of the assembly under steady state conditions is shown in Fig. 7.1.

7.1.2 Temperature Distribution in FBTR Storage Fuel Subassembly (SA) during postulated Complete Flow Blockage Scenario

As part of regulatory review for renewal of licence for operation of FBTR, an independent verification study was carried out to assess the effects of complete flow blockage scenario in a storage fuel subassembly. A detailed CFD study was undertaken to determine whether the clad temperature and sodium temperature exceed the permissible limits under the postulated scenario. The temperature distribution in storage SA is depicted in Fig.7.2. It was found that the peak clad temperature remains within the limiting peak clad temperature of 823K. The maximum sodium temperature is also well within the boiling point temperature of 1153K.

Fig. 7.1: Temperature Contour (K) at top (z=160 mm) of ISZ100 with 74% Flow Blockage

Chapter 7SAFETY ANALYSIS AND

RESEARCH


7.1.3 Conceptual Design for Passive Containment Cooling with Application to PHWRs

A conceptual model of PCCS for 700 MWe PHWR containment [Fig 7.3(a)], consisting of parallel heat removal loops has been developed. Preliminary studies have indicated that a 6-loop PCCS would be able to remove the decay heat load.

Performance assessment of the conceptual design was carried out by developing and applying a two-phase natural circulation computer code to simulate flow in non-uniform diameter, parallel channels. The code was validated against experimental data available in literature. The influence of various geometrical and operating parameters on the natural circulation flow characteristics has been studied [see Fig 7.3 (b)]. This study has provided inputs for setting up a scaled down PCCS experimental facility at SRI.

7.1.4 Thermal Hydraulic Analysis of Containment Filtered Venting System of TAPS-3&4

As part of independent assessment, analysis of TAPS-3&4 CFVS was carried out using RELAP5 for the postulated event of LOCA with loss of ECCS and loss of moderator cooling with SAMG action of water injection into the calandria vault. Depressurisation of the containment with operation of the CFVS was analysed during the mission time of 7 days. Pressure and vapour pressure variation in containment and scrubber tank during CFVS operation is shown in Fig. 7.4. It was observed that the scrubber tank inventory reaches steady level due to continuous steam condensation and removal

Figure 7.2 : Temperature Contour (K) at Mid-plane of Storage Subassembly

Fig 7.3(a): Conceptual PCCS Model Fig 7.3(b): Effect of Heat Transfer Coefficient on PCCS Mass Flow Rate


through overflow line. The liquid to gas ratio in the venturi was found to be more than 7 which is suitable for effective decontamination of fission products.

7.1.5 Annulus Gas Monitoring System (AGMS) Analysis

7.1.5.1 Simulation of Injection Test Conducted in KAPS-2 under Cold Shutdown Conditions

The original purge mode operation of AGMS in KAPS-2 is being converted to recirculation mode after EMCCR. To assess the performance and sensitivity of the upgraded system, a moisture injection test (using bubbler) was conducted under cold shutdown conditions. The system was in recirculation mode with a flow rate of 30 m3/h and the initial dew point was -23.5°C. Moisture was injected for 3 hours and a total 150g of water was carried away by the gas through a bubbler. Thermal hydraulic simulations using RELAP5 computer code were carried out for moisture injection test conducted at KAPS-2 using two different AGMS models namely KAPS-2 and standard 220MWe PHWR AGMS models. The dew point temperature prediction of KAPS-2 and Standard 220 MWe

PHWR AGMS models for steam and water injection rate of 50 g/h is as shown in Figs. 7.5(a) & 7.5(b) respectively. It was concluded from the analysis that 50g/h water injection is in fair agreement with the experimental data. Additional analysis to determine the actual moisture content in the system was also carried out and it was found that moisture injection through bubbler was equivalent to that obtained from 4g/h (as shown in Fig. 7.5(a)) of steam injection into the system.

Fig. 7.4: Pressure and Steam Partial Pressure variation in Containment and at CFVS Exit

(a) KAPS AGMS model (b) Standard 220 MWe AGMS model

Fig. 7.5: Comparison of Dew Point Temperature Analysis with Experiment


7.1.5.2 AGMS Response after Removal of Two Pressure Tubes (PTs) in 220 MWe PHWR

Two Pressure Tubes (PT) from RAPS-3&4 and KSG-1&2, were removed from the high flux regions, for material surveillance purpose. A study was carried out to estimate AGMS response time for a postulated leak from PT for the changed and the initial configuration. Two different postulated leak rates (30 g/h & 300 g/h) from the PT under recirculation and purge mode of operations were considered in the analysis. The comparison of variation in dew points under recirculation and purge mode of operations for both the configurations are shown in Figs. 7.6(a) & (b) respectively. It was concluded from the analysis that the difference in dew point at sensor location for initial and changed configuration was not significant.

7.2 SEVERE ACCIDENT STUDIES

7.2.1 Assessment of Accident Progression in MAPS / RAPS Reactors

The older generation PHWRs (MAPS/ RAPS) have many components and safety systems different from current generation PHWRs and pose several challenges in assessment of postulated Design Extension Conditions (DEC). It is seen

that available safety analysis codes (RELAP/SCDAP, ASTEC, etc.) do not have models for all phenomena relevant to DEC in older generation PHWRs. Hence, a in-house programme called Severe Accident Analysis Programme (SAAP), which was earlier used for in-channel retention studies, was developed further for carrying out in-vessel retention studies in MAPS/ RAPS.

Numerical studies were carried out for postulated event initiated by LOCA and failure of moderator pumps. The analysis was carried out up to coolant channel disassembly and failure of dump port. The results were refined based on outcome of code benchmarking exercise among participants from BARC, NPCIL and AERB. The code was being further developed to study the progression of the event beyond in-Calandria phase of the event.

7.2.2SAMGVerificationModeratorExpulsionAnalysis

Addition of water to calandria vessel is one of the prescribed SAMG action for Indian PHWRs. The water injection requirement as a part of this action depends not only on the heat coming into the moderator but also on the extent of moderator expulsion subsequent to OPRD rupture. A thermal-

(a) Recirculation Mode (b) Purge Mode

Fig. 7.6: Dew Point variation for 30 g/h and 300 g/h Leak Rate from PT


hydraulic study was carried out to evaluate the extent of moderator expulsion upon OPRD rupture following a DEC scenario using RELAP5 computer code. Parametric study with respect to moderator temperature, flow discharge coefficient and thermal non-equilibrium constant (Cne) was carried out to study the influence of these parameters on moderator expulsion. Variation in moderator inventory with different moderator temperatures at the time of rupture is as shown in Fig. 7.7(a) and for different values of Cne for fixed moderator temperature is shown in Fig. 7.7(b). It was concluded from the analysis that the moderator temperature at the time of OPRD rupture and Cne has a significant influence on the moderator expulsion from calandria vessel.

7.2.3 Hydrogen Distribution and Mitigation Analysis with PCRD for KAPS 700 PHWR Containment using 3D CFD tool

CFD analysis was carried out with the objective to find out the local hydrogen and steam distribution in containment in presence of hydrogen mitigation devices. A 3D structured mesh was generated for the containment. The conjugate heat transfer approach was used for wall and fluid

interface. Accident scenario considered for the simulation was LOCA with Loss of Emergency Core Cooling System (LOECCS) coincident with loss of moderator cooling. This analysis covered entire containment with condensation phenomenon.

7.2.4 CFD Simulation of MAPS Calandria Temperature Analysis

As part of assessment of safety of old PHWRs, a benchmark analysis was carried out using CFD tool FLEUNT to understand effect of radiation heat transfer on calandria temperature. A detailed 2D structured meshed geometric model considering all 306 channels was developed. The equivalent PT-CT, calandria, calandria vault were modelled with appropriate boundary and initial conditions. Conjugate heat transfer approach for wall with thickness was also considered. Discrete Ordinates (DO) model and P1 radiation theory models were applied. It was confirmed that the radiation was predominant mode of heat transfer from channels, which contributed more than 90% of overall heat transfer. The outcome of this analysis was being used suitably in the integrated severe accident analysis of LOCA with loss of moderator circulation. Figure-7.8(a), (b) and (c)

(a) Variation with temperature (b) Variation with Non-equilibrium constant (Cne)

Fig. 7.7: Moderator Inventory with Variation in Temperature and Non-equilibrium Constant


indicate full 2D structure mesh (74000 elements) of the geometry.

7.3 SAFETY ANALYSIS CODE DEVELOPMENT

7.3.1 Contribution to DAE Computer Code for Severe Accident Simulation

DAE computer code for Severe Accident Simulation (PRABHAVINI) is under development and AERB is contributing models for various safety

systems to this generic code. As part of this, a point model for Passive Catalytic Recombiner Device (PCRD) was developed as a separate module. This module gives user choice of reaction chemistry models (single step, multi-step or aggregate model) and computed the rate of reaction, consumption and production of various gaseous species and enthalpy addition in various containment compartment. This module was validated and integrated in PRABHAVINI.

(a) Computational Domain

(b) Detailed view of Mesh (c) Closer view of Mesh

Fig. 7.8: Computational Domain


7.3.2 Development of in-house Nuclear Plant Analyser Code (NuPAC)

As a part of safety analysis code development programme, 1-D transient thermal hydraulic code, NuPAC, was developed. Transient thermal-hydraulic, two-phase phenomena were calculated from formulations of one-dimensional, homogeneous, equilibrium conservation equations for mass, momentum and energy. Heat structures were modelled using a transient one-dimensional heat conduction solution that was coupled to the fluid through heat transfer relations. NuPAC code is based on finite volume method and employs a semi-implicit numerical scheme for solution of governing equations. The water-steam property package based on IAPWS95 standard was also developed and integrated with NuPAC code. The NuPAC code has the capability to simulate closed loop, open loop and parallel flow systems. An Inter-code comparison with RELAP5 for single phase natural circulation system was completed and results were in good agreement as shown in Fig. 7.9.

7.3.3 Development of Source Term Evaluation Tool

A GUI based software tool was developed for estimating the ‘Source Term’ as per ‘Radiological

Impact Assessment (RIA)’ guidelines of AERB. The tool is capable of estimating the source term for different reactors considering different postulated accident scenarios of selected plants. The drop down menu to select different plants and scenarios is provided in the tool. The graphical and tabular representations of various thermal hydraulic input and fission product release are also available. The user interface of the tool is shown in Fig. 7.10. The source term evaluation tool was successfully and effectively utilised in the off-site emergency exercise conducted at RAPS on December 15, 2018 for the estimation of source term.

7.3.4 Development of Code Systems for Coupled Thermal Hydraulics-Neutronics Analysis of Large Core Nuclear Power Reactors

A computer code system was developed to perform coupled Thermal Hydraulics-Neutronics analysis of large loosely coupled power reactors for system induced transients involving core neutronic asymmetry. The model was coupled with system Thermal Hydraulic code RELAP5 and was validated against a highly asymmetric fast neutronic transient benchmark representing LOCA in a large PHWR. The model was used to simulate asymmetric power rise in a VVER-1000 for start-up of an idle Main Coolant Pump

(a) Mass flow rate (b) Pressure

Fig. 7.9: Inter-code Comparison of NuPAC and RELAP5 Codes


(MCP) when other three MCPs are in operation. The data from a reference VVER was used to demonstrate the capability of the model to predict core asymmetricity coupled with system thermal-hydraulics. The model successfully captured the phenomena of asymmetric power rise of the affected loop. This work supplemented AERB’s in-house capability to perform independent verification analysis involving coupled core neutronics and system thermal-hydraulics modelling for a number of transients in the existing and upcoming large NPPs.

7.4 RADIOLOGICAL ASSESSMENT AND ENVIRONMENTAL SAFETY STUDIES

7.4.1 Dispersion and Dose Calculation for Protective Actions during an Offsite Emergency

To carry out independent monitoring during an emergency condition, AERB carries

out assessment of the situation. In its efforts to strengthen the in-house capability in addition to monitoring the off-site emergency exercises that are being conducted at different NPP sites, also carries out evaluation of source term and estimation of projected dose in the public domain. Towards this, necessary activities for source term estimation and dose calculation were carried out during the off-site emergency exercise conducted at Rajasthan Atomic Power Station (RAPS) on December 15, 2018. The postulated accident scenario was LOCA along with loss of Emergency Core Cooling System (ECCS), loss of moderator cooling and containment Embedded Plate (EP) failure. By using the calculated source term for the given scenario, atmospheric dispersion modelling was carried out for important radionuclides such as I131, Cs137, Cs134, Xe133 and Te132. Out of the five radionuclides studied I131 (½ life ~ 8 days) was found to be the major contributor to the radiation

Fig. 7.10: Source Term Evaluation Tool Interface


dose. The studies also indicated availability of adequate time for appropriate counter measures.

7.4.2 Estimation of Radionuclide Release from PHTS to Containment for PHWR-700 DEC-B Scenario

Analysis for postulated initiating event for design extension condition (DEC) with core melt scenario for PHWR-700 reactor was carried out based on the radiological impact assessment guidelines. The detailed thermal hydraulic analysis of LOCA without ECCS and moderator circulation was carried out using code RELAP5 for the duration of around 7 days. In-core inventories of 82 radionuclides for the equilibrium core were estimated through fuel burn-up analysis. Overall integration of all calculations was carried out by in-house developed computer code SASTAT (Severe accident source term analysis tool). Steady state gap inventory of radionuclides due to the normal power operation before accident took place was estimated. The release of radionuclides from fuel rod to coolant was estimated based on fuel failure criteria given in the guideline. The temperature dependent transient release of radionuclides to coolant was predicted. The impact of radiation decay of parent nuclides and build-up of daughter nuclides in the fuel matrix on the release to containment was analysed. The contribution of the short-lived radionuclides was found to be reduced significantly due to implementation of radioactive decay chain in the fuel matrix. Estimation of the radioactivity that would be released to the containment from the primary heat transport system (PHTS) was performed.

7.4.3 Mapping of Condenser Coolant Discharge along Kudankulam Sea - By processing Satellite Thermal Data

A study was taken up to assess the thermal discharge pattern due to the condenser coolant discharge into the sea around KKNPP plant. The study on the pattern of dispersion as baseline

data is important as unit 3&4 and 5&6 are under construction. The spatial and temporal characteristics of the thermal plume spread with respect to seasons was studied with satellite Thermal Infra-Red (Landsat 7 band 6 B2, Landsat 8, band 11) data. The processed satellite imagery data showed that the difference between the intake and the mixing zone at receiving body was within the limit prescribed by MoEF.

7.4.4 Nuclear Emergency Management Information System for Rawatbhata and Narora sites

A user-friendly query based information retrieval system, NEMIS (Nuclear Emergency Management Information System) for off-site emergency management during nuclear accidents was developed using Visual Basic code system and integrated with GIS software. As part of this work, site specific geo-spatial database for emergency planning was generated for Rawatbhata and Narora sites (Fig. 7.11). This GIS based system was developed to provide real time information such as possible affected area, population to be advised on protective actions, evacuation routes, transport requirement, nearest hospitals, primary health centres etc., for proper emergency preparedness and management.

7.4.5 Preparation of Land use/Land cover map for NPP sites

A study was undertaken to generate geo-spatial database on environmental baseline data for all nuclear plant sites. As part of this study, a land use / land cover map was prepared for Kudankulam site. The land was categorised into various land use / land cover type by employing LANDSAT 8 based on the NRSC classification manual. The major type of land use classes comprised in the study area included evergreen forest, deciduous forest, scrub forest, forest plantations, small patches of agricultural lands, wastelands, coastal plantation, water bodies and settlements. This database would


be used as base data for comparing with the post commissioning data to assess possible stress on the environment due to the plant activity.

7.5 PROBABILISTIC SAFETY STUDIES

7.5.1 Human Reliability Analysis

During review of PSA reports submitted by utilities, it is observed that different human reliability models are adopted to quantify human errors and there is no standard methodology adopted. To identify a suitable methodology for human reliability analysis, AERB PSA committee for Nuclear Facilities carried out an exercise. Detail of various human reliability models was deliberated and based on the merits and demerits of various methods, it was decided to follow human

cognitive reliability model for diagnosis errors in combination with Accident Sequence Evaluation Programme for execution errors.

7.6 EXPERIMENTAL STUDIES

7.6.1 Cable Fire Studies

Experiments on power and control cable fires were conducted in the Compartment Fire Test Facility (CFTF). The focus was on performing confirmatory tests on behaviour of aged cables. Test samples of XLPE and FRLS PVC cables were prepared and tested in vertical-tray flame test configuration, as per IEC 60332 and IEEE 1202, in uncoated condition as well as with intumescent paint coating. Major parameters of interest were spontaneous ignition, swelling, decomposition,

(a) Road Map of Narora site (b) Rallying post locations of Narora site

(c) Road Map of Rawatbhata site (d) Rallying post locations of Rawatbhata site

Fig. 7.11 Geospatial Database for Off-site Emergency Management


flame propagation cable damage, mass loss and heat release rates, smoke generation rate etc. Surface temperature along the cable surface was also obtained using thermocouples and thermal imaging. Fig.7.12 shows a cable fire experiment and infrared images taken during the heating and cooling period. These tests have demonstrated the effectiveness of intumescent paint coating, even in aged cables.

7.6.2 Hydrogen Mitigation Facility

As part of ongoing severe accident mitigation studies, an experimental facility “Hydrogen Mitigation Facility (HYMIF)” was commissioned

in the SRI Engineering Hall (Fig. 7.13(a)). The objective of the facility is to obtain important safety parameters for catalytic recombiners. Parametric experimental studies were in progress in association with BARC and IGCAR. Hydrogen-air mixture at ambient temperature, 1-4% H2 concentration (V/V) and flow velocity in the range of 0.2-0.6 m/s is supplied at the inlet to the PAR. Experimental data on temperature transients and gas concentration were obtained and reaction rates were computed (Fig. 7.13(b)). Subsequently, catalyst coupons with Pt on SS mesh were prepared and experimental studies with these catalyst coupons was in progress.

Fig.7.12: (i) Snapshot of a Cable Fire Experiment; and (ii) Infrared Images of Cable Surface [a) 30s b) 1 min c) 6 min d) 21 min (cooling period)]

Fig. 7.13 (a) Isometric view of HYMIF; and (b) Reaction Rate obtained using two vertically placed Pt coated cordierite plates in Catalyst Chamber


7.6.3 Commissioning of Core Melt Retention Facility (COMREF) and Material Characterisation Studies

With an objective to demonstrate in-vessel corium retention capability of calandria vessel during a core melt accident, a Core Melt Retention Facility (COMREF) was being set-up at SRI, Kalpakkam. The facility simulates a hypothetical severe accident condition in which molten corium is inside the calandria vessel and outside surface is cooled by calandria vault (CV) water. For the facility, test vessel and water vault were fabricated to simulate the calandria vessel and calandria vault. A trial run of integrated system was completed (Fig. 7.14) and experimental studies were in progress.

Also, for hypothetical severe accident condition in which even calandria vault water is not available, temperature of calandria vessel will rise to very high temperature. An experimental study was also initiated to study the high temperature tensile and creep properties of calandria vessel material and to develop a constitutive model based on damage mechanics.

7.6.4 AGMS and Coolant Channel Heat-up Facility

An experimental facility was being set-up at SRI-AERB for investigating coolant channel heat-up and annulus gas monitoring system related safety issues [Fig.7.15]. There are nine PT-CT

assemblies and each is provided with tubular heater. The power input to individual heaters is controlled from a panel using Solid State Relay based control system. Safety Report of the AGMS test facility was submitted to IGCAR Safety Committee for review.

7.6.5 Water and Steam Interaction Facility (WASIF)

An experimental facility is being setup within the high bay of SRI engineering hall in collaboration with BARC to investigate various forms of Direct Contact Condensation (DCC) phenomena [Fig.7.16]. Erection of equipment and installation of instrumentation system for this facility were in an advanced stage of completion. A separate control room was constructed to house

Fig 7.14: COMREF Test set-up during Trial Run Fig.7.15: AGMS Test Facility

Fig.7.16: WASIF Facility


the main control panel. The functionality test of the main control panel was completed recently. A revised safety report of WASIF was submitted after incorporating the recommendations and suggestions of IGCAR Safety Committee.

7.6.6 In-house Experimental Investigations on Corrosion Aspects of Zr-2.5%Nb Material

Subsequent to the pressure tube failure incident at KAPS-1 nuclear reactor, in-house R&D studies on the corrosion aspects of Zr-2.5%Nb have been taken up to investigate the causes of corrosion. Parameters such as elemental composition,

phase and morphology as well as roughness measurements of the corrosion signatures were critically measured for each experiment. Scanning Electron Microscope (SEM) micrographs of the exposed specimens indicated the disturbance to the oxide layer including peeling off oxide film, crack and formation of nodular features (Fig. 7.17). X-ray Diffraction (XRD) determinations showed the presence of monoclinic phase clearly indicating the presence of ZrO2 in the nodules. Elemental mapping using Electron Dispersive Spectroscopy (EDS) indicated mainly the presence of Zr, Nb and oxygen (Fig. 7.18).

(a) Fresh (b) Exposed to Formic Acid

Fig. 7.17 Image of Zr-2.5% Nb Specimen

Fig. 7.18 (a)SEM Micrograph of Zr-2.5%Nb Specimen (after formic acid exposure) and (b) EDS Spectra for Selected Corrosion Spot


Experimental observations also revealed that ethylene, if present, as an impurity in annular gas CO2, gets oxidised to form corrosive intermediates such as formic acid. The formation of intermediates was also confirmed through theoretical modelling at SRI by using software, FACTSAGE. Once formic acid is formed, the ZrO2 protective layer gets disturbed, becomes porous, gets peeled off and forms weak spots for other impurities such as hydrogen, moisture etc. to enter and cause further damage to Zr-2.5%Nb substrate. Further, it could be established that the ionising radiation catalyses the oxidation through the formation of free radical intermediates. Diffusion of hydrogen up a stress gradient or down a thermal gradient is responsible for stress-induced reorientation of hydride precipitates, delayed hydride cracking and formation of hydride blisters at cold spots on the Zr-2.5%Nb tube. Experiments were also carried out using the other possible impurities in the annulus gas such as methane etc. The in-house experimental facility is capable of carrying out similar corrosion experiments with other material samples also.

7.6.7 Investigation of Optically Stimulated Luminescence (OSL) Response of Various Common Materials for Retrospective Dosimetry

In the event of a radiological accident, prompt assessment of the dose levels received by radiation workers and by the general population is important. As general public do not wear any kind of personal dosimeters, the dose estimations for these individuals must be performed using alternative methods. One of the important tools for such accidental dosimetry is the retrospective dosimetry, which makes use of luminescence property of materials available in the contaminated area to obtain dose information. In this regard, investigation was carried out to determine Optically Stimulated Luminescence (OSL) response of various materials such as common salt, porcelain

substrate, bricks & other building materials, watch glass and electronic components found in mobile phones, intercoms and computers etc. The samples were irradiated in gamma chambers and analysed for OSL property by stimulating with blue and green LEDs using a TL/OSL reader. It was inferred from the study that the ceramic based electronic components and building materials showed better OSL response. Further investigations were in progress.

7.7 REACTOR PHYSICS STUDIES

7.7.1 VVER Regulator’s Forum Benchmark Exercise on Coupled Neutronics and Thermal Hydraulics

A bilateral benchmark exercise on a Coupled Neutronics and Thermal Hydraulics was developed jointly by Russia and India as one of the activities of the Working Group on Reactor Physics (RPWG) of VVER Regulator’s Forum. As a first phase of the benchmark exercise, lattice physics and core neutronics analyses were carried out using indigenous code systems for VVER-1000 core comprising of gadolinium-based fuel assemblies. Important core physics parameters were evaluated and compared with the results of Russian code systems. Comparison demonstrated the capability of indigenous computer code systems at par with the international codes.

7.7.2 A Benchmark on Coupled Code Systems for Plant Dynamics Studies of Indian NPPs

Transient response evaluation of NPPs has been one of the important dimension of nuclear reactor safety. As a conventional practice, until recent years most of the safety analyses have been successfully performed using thermal-hydraulics system codes. However, these codes have limitations in situations where complex feedback exists between core neutron kinetics and thermal-hydraulics or when core and/or system asymmetric


phenomena are involved. Availability of strong computing resources now-a-days rendered the flexibility to use the complex coupled computer codes simulating multi-physics phenomenon. This has made direct modelling of the interaction between the 3D neutron kinetics and the thermal-hydraulics phenomena feasible. This approach consists of incorporating three-dimensional kinetic modelling of reactor core into thermal-hydraulic system codes. One of the important benefits of such analysis approach is relaxation of safety margins without compromising the NPP safety, allowing higher operating power and extended fuel cycles. Many initiatives have been taken up worldwide as well as in India also to couple core 3D kinetics codes with system codes. Such coupled code system require intensive validation to establish its credibility as a best estimate analysis tool. However, availability of data on these international benchmark problems is limited. To provide a reference benchmark for Indian coupled code systems, used for best estimate analysis of NPPs, AERB took an initiative to define a detailed multi physics benchmark problem for coupled codes. Different teams from various DAE units are participating in this round robin exercise. The exercise would be completed in two phases. The first phase focuses on stand-alone neutronics and stand-alone thermal hydraulics. Second phase exercise will be full-fledged integrated 3D neutron kinetics with detailed system thermal hydraulic analysis of IPHWR-540 core during a defined transient scenario.

7.7.3 Steady State Reactor Physics Analysis of 700 MWe PHWR

Steady state reactor physics analysis of 700 MWe PHWR was carried out in-house by independent computer codes and cross-section data. The initial core loading patterns of the reactor, comprising of 28 thorium/ 200 DU/ 248 DDU bundles were analysed for evaluating intial

core excess reactivity, worth of reactivity devices and various reactivity coefficients. Hot Full Power (HFP) condition with the equilibrium value of xenon and cold condition were simulated with the nominal positions of reactivity devices. Comparsion of the design calculations and independent simulation results showed good agreement for all values.

7.7.4IndependentVerificationofReloadSafetyEvaluations for Kudankulam VVER-1000 Reactors

Indigenous code systems have been extensively used for independent verification of design calculations on reload safety evaluations for various operating cycles of Units-1&2 of Kudankulam VVER-1000 reactors and for analysis of the associated low power physics experiments. Review calculations have been carried out for Cycle-4 of Unit-1 & Cycle-2 of Unit-2 KKNPP. Important core safety parameters were calculated and compared with the design report. Comparison showed the expected trends and good agreement with the design calculations and also with the measurements.

7.7.5 Development of Space Time Kinetics Module for Hypothetical Core Disruptive Accident Analysis of Fast Reactor

As a part of fast reactor safety analysis, a space-time kinetics computational model was developed by solving the time dependent neutron diffusion equation in two-dimensional cylindrical geometry with advanced numerical schemes. Two benchmark problems, e.g. D2O reactor super critical transient and fast reactor super prompt critical transient were analysed by the developed computational module. The results of the analysis were in good agreement with the benchmark results. The space-time kinetics module was coupled with hydrodynamics module to analyse disassembly phase of hypothetical core disruptive accidents in fast reactors.


7.8 STRUCTURAL ANALYSIS AND MATERIAL STUDIES

7.8.1 Evaluation of Nodular Oxide Induced StressfieldinZirconiumAlloyTubes

Oxide nodules formed due to nodular corrosion of the Pressure Tube (PT) induces high stress at the interface of oxide layer and base material. The induced stress fields form favourable sites for hydrogen accumulation which may later lead to Delayed Hydride Cracking (DHC) at these locations. Considering the important role of induced stress in DHC, a finite element analysis was carried out to evaluate induced stress in PT due to isolated and overlapped nodules using ABAQUS. Axisymmetric modelling was used to simulate spherical nodules having various sizes and length to depth ratio (l/d). Induced stress and strain fields were evaluated at nodule bottom point and middle point as shown in Fig. 7.19. It was observed that the induced stress field in the overlapped nodules were higher than isolated nodules. A parametric study was also performed to study the effect of granular ZrO2 properties, nodule sizes, aspect ratios etc. on the induced stress field in PT. Fig 7.19 shows the contour plot of hoop stress for overlapped nodules.

7.8.2 3-D Simulation of Roll Expansion Joint of Pressure Tube to End Fitting in PHWR

In PHWR, Zr-2.5%Nb PT and stainless-steel end fittings are roll expansion joined by localised radial expansion of the PT using a rotating roller. Inherent with this roll expansion is development of tensile residual stress field within the tube with maximum value in the transition region from expanded to unexpanded zone of the tube. The residual stress field in transition region is an important input for assessment of fitness for service of pressure tubes in PHWR. To better understand the roll joint performance, a 3-D numerical model of the PT to end fitting roll expansion joint was developed using explicit dynamic Finite Element (FE) procedure in ABAQUS. The numerical modelling of the rolled joint is complex because of 3-D nature of the problem involving large strain plastic deformation with cyclic loading unloading and moving contact. Taking into account all these complexities, non-linear analysis of the rolled joint was performed and complete displacement and stress field was obtained in the rolled as well as transition region of PT (Fig.7.20).

The complete stress field of the rolled joint offered immense possibilities of assessing the effect

Fig.7.20: Residual Hoop Stress distribution

in Rolled JointFig. 7.19: Contour Plot of Hoop Stress for Overlapped Nodules


of various parameters on the roll joints which was difficult to obtain from other method like testing.

7.9 SAFETY STUDIES TO SUPPORT REVIEW AND ASSESSMENT

7.9.1 Soil Structure Interaction Analysis of Combined Pile Raft System

4 no. of 700 MWe PHWR based NPPs are proposed to be constructed at Gorakhpur Haryana (GHAVP) site. GHAVP is an alluvial site with depth to bed rock estimated to be around 350 m. Soil structure interaction (SSI) is a major consideration for heavy structures founded on such soil. Combined pile raft foundation system (CPRF) was proposed for the GHAVP site. Towards this, a series of dynamic centrifuge tests on a piled raft system was selected from literature to validate ACS SASSI software program (Fig. 7.21). Different pile modelling techniques were studied and compared. Responses of the pile raft system and the bending moment of the pile were compared for a medium level earthquake (Fig. 7.22). Based on the study, it was concluded that pile modelled as solid element was able to represent realistic behaviour of pile raft system as observed in centrifuge tests.

missile induced vibrations from the impacted civil structure to floors and walls, which are outside the impacted area was investigated. A reinforced concrete target was impacted chronologically by three soft missiles having mass of 50 kg and velocities of values should be changed to 91.8 m/s, 93.5 m/s and 167 m/s respectively. Numerical simulation was carried out using non-linear transient dynamic analysis with explicit solution technique of ABAQUS software. In this ‘study’, structural responses of different parts of target and two pseudo equipment, which were connected with rear side of target wall using bolts and welds were simulated. After analysing the target for one impact, the damaged target was carried over to next level of impact to simulate exact condition of experiments. Figure 7.23 (a), (b) and (c) below shows the target and missile before impact, the comparison of deformed missile simulation vs experiment after 1st impact (91.8 m/s) and the comparison of deformed missile simulation vs experiment after 3rd impact (167 m/s) respectively. Major responses obtained from analysis compared well with the experimental findings.

7.9.3 Benchmarking Exercise for Capacity Assessment of Shear Walls

As part of benchmarking exercise for capacity assessment of shear walls (CASH) organised by OECD-NEA, wherein AERB is a participant, finite element modelling of full

Fig. 7.21 Modelling Approach for Pile Foundations in SASSI

7.9.2 IRIS-Phase-III Exercise Conducted by OECD-NEA

As part of the international round robin exercise on ‘Improving the Robustness assessments methodologies of structures Impacted by missile S-Phase-III’ (IRIS-Phase-3), transmission of

Fig.7.22 Response Spectra at top of Raft for Pile Modelled as Solid, Beam and Beam with Rigid Links


scale shear walls was undertaken. As required, response spectrum, nonlinear static pushover and nonlinear time history analysis for set of input earthquake time histories were carried out. Failure mode of shear walls were observed to be consistent across different methods viz.

nonlinear static pushover and time history analysis. Results of pushover and nonlinear time history analysis in terms of damage pattern for regular and irregular shear walls at different levels of earthquake are shown in Figure-7.24 and Figure-7.25 respectively.

(a) Numerical Model of Test Set-up (b) Comparison of Deformed Missile after Impact of 91.8 m/s

(c) Comparison of Deformed Missile after Impact of 167 m/s

Fig. 7.23 Numerical Model of Test Set-up and Comparison of Deformed Missile after Impact of 91.8 m/s and 167 m/s

(a) Regular Shear wall (b) Irregular shear wall

Fig. 7.24: Damage Levels from Pushover Analysis at different Levels of Earthquake


7.10 AERB FUNDED SAFETY RESEARCH PROGRAMME

AERB promotes and funds research in radiation safety and industrial safety as part of its programme. AERB Committee for Safety Research Programmes (CSRP) frames guidelines and evaluates, recommends grants for research projects and monitor their progress periodically. During this period, CSRP recommended five new

Table 7.1 : New Research Projects Approved

S. No. Project Title Principal Investigator Organisation

1. Study of fundamental heat transfer characteristics in the presence of non-condensable for designing long term passive heat removal system for containment

Dr. Arun Kumar Sridharan IIT Bombay, Mumbai

2. Phytoremediation of radioactive elements (Cesium and Strontium) from contaminated soil and water

Dr. N. K. Dhal CSIR- IMMT, Bhubaneswar

3. Determination of anisotropic elastic constants and anisotropic yield parameters for Zr 2.5% Nb pressure tubes

Dr. Avijit Kumar Metya CSIR-NML, Jamshedpur

4. Numerical Crack growth studies in hydrided pressure tube of PHWR

Dr. Indra Vir Singh IIT, Roorkee

5. Low pressure nanofiltration for removal of monovalent and bivalent salts from leached liquor during alkaline Uranium ore processing

Dr. Sirshendu De IIT, Kharagpur

(a) Regular Shear wall (b) Irregular shear wall

Fig.7.25: Damage levels from Nonlinear Time History Analysis at different Levels of Earthquake

projects. It also approved the renewal of four ongoing projects. The details are given in Tables 7.1 and 7.2.

AERB also provides financial assistance to Universities, Research Institutions and Professional Associations for holding symposia and conferences on the subjects of interest to AERB. During this period, financial assistance was provided to 43 Seminars, Symposia and Conferences.


• DemonstrationofCSRPStudyResults

CSRP had sanctioned research project titled ‘Fabrication of Nano oxide based Sensor on Stabilised Nano Zirconia for Detection of Hydrogen Sulfide’ to Rajalakshmi Engineering College (REC), Chennai. Under this project, indigenous nano based H2S sensor was developed and tested

Table 7.2: Research Projects Renewed

S. No.

Project Title Principal Investigator Organisation

1. A study on the role of colloids in the transport of radionuclides in water

Dr. S. Chidambaram Annamalai University, Tamilnadu

2. Synthesis of Chitosan based Ploy Electrolyte Ultrafiltration Membrane for the remediation of Cesium from aqueous media

Dr. M. Dharmendira Kumar

Anna University, Chennai

3. Studies on environmental radioactivity levels in and around Chitrialuranium mineralised areas of Nalgonda District, Telangana State

Dr. Ch. Gopal Reddy Osmania University,Hyderabad

4. Utilisation of Steel Slag materials in concrete for radiation shielding studies

Dr. V. Ganesh Kumar Satyabhama University, Chennai

in the laboratory. Performance of the sensor was further tested in the Chemical laboratory of Heavy Water Plant, Manuguru, at various concentrations of H2S gas during May 2018 by a team of AERB and REC officers. The sensor was found detecting the H2S gas at 150oC. The demonstration showed an indirect indication of sensor behaviour with and without exposure to H2S.

AERB, REC and HWB Officers at HWP, Manuguru during the H2 S sensor testing

chapter 7 safety analysis and research · 78 aerb annual report - 2018 indicate full 2d structure...

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