Inlet Header Break Research Articles

BEPU analysis of a CANDU LBLOCA RD-14M experiment using RELAP/SCDAPSIM

A key element of the safety analysis is Loss of Coolant Analysis (LOCA) which must be performed using system thermal-hydraulic codes. These codes are extensively validated against separate effect and integral experiments.RELAP/SCDAPSIM is one such code that may be used to predict LBLOCA response in a CANDU reactor. The RD-14M experiment selected for the Best Estimate Plus Uncertainty study is a 44 mm (22.7%) inlet header break test with no Emergency Coolant Injection. This work has two objectives first is to simulate pipe break with RELAP and compare these results to those available from experiment and from comparable TRACE calculations. The second objective is to quantify uncertainty in the fuel element sheath (FES) temperature arising from model coefficient as well as input parameter uncertainties using Integrated Uncertainty Analysis package.RELAP calculated results are found to be in good agreement with those of TRACE and with those of experiments. The base case maximum FES temperature is 335.5 °C while that of 95% confidence 95th percentile is 407.41 °C for the first order Wilk's formula. The experimental measurements fall within the predicted band and the trends and sensitivities are similar to those reported for the TRACE code.

Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—II: Full-Core Analyses

The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Earlier lattice physics and reactor core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic (T-H) studies are required to confirm that these advanced fuels will have adequate T-H safety margins. Preliminary system T-H transient simulations have been carried out for a 700-MW(electric)–class PT-HWR in a postulated loss-of-coolant accident (LOCA) using the CATHENA code. One purpose of this work was to demonstrate that such simulations of a PT-HWR filled entirely with advanced fuels could be set up and executed successfully in a CATHENA transient simulation model. The other purpose was to evaluate the peak sheath and peak fuel centerline temperatures during a LOCA to perform an analysis that compares the relative performance of each of the proposed advanced fuels. System T-H simulations with CATHENA were performed to model a postulated LOCA event with a 20% inlet header break in a typical 700-MW(electric)–class PT-HWR using two types of advanced uranium-based and thorium-based fuel bundles in modified 37-element and 35-element geometries. Calculations were also performed for a PT-HWR using conventional natural uranium fuel in 37-element fuel bundles for comparison. In the event of a LOCA, there is a drop in the primary circuit pressure. It is assumed that there is a 2-s delay between the signal of the low primary pressure and the tripping of the reactor. When the reactor trips, the shutdown rods are inserted. The reactor trip is followed by the activation of the emergency core cooling system, which occurs 30 s after the LOCA starts, with a trip signal on the boiler crash cooling. Simulation results for the LOCA demonstrated that the peak fuel centerline temperatures (ranging from 1822°C to 2183°C) were several hundred degrees below the expected melting point of UO2 (~2865°C). Simulations also demonstrated that the peak sheath temperatures for the advanced fuel concepts ranged from 1177°C to 1204°C, which are lower than that with conventional NU fuel in 37-element fuel bundles. Thus, the system T-H analysis of the relative results provides confidence in the proposed advanced uranium-based and thorium-based fuel concepts for potential use in PT-HWRs.

Numerical assessment of LOCA experiments of RD-14M using MARS-KS code

The code assessment on the design basis events has been of fundamental importance to the safety analyses of the nuclear power plants. When a system analysis code is released to the new version by improving the existing models and correcting the found errors, it is essential to make sure that the updated code can predict and simulate reasonably the important phenomena possibly observed during the transient. This work has focused on strengthening the predictability and reliability of the system analysis code by benchmarking the invaluable loss-of-coolant-accident (LOCA) data obtained from an integral effect test facility (RD-14M). In this study, the verification tests of the inlet-header break simulating the LOCA of the CANDU-6 plants were selected as a benchmark case for MARS-KS code evaluation. First, a proper analysis model to benchmark the RD-14M experiment was envisioned into the MARS-KS input model and its steady-state conditions were achieved successfully. Subsequently, through the transient analysis experimental data of void fraction, header pressure, differential pressure between headers, and sheath temperature were compared with the code results. The code input models predicted the experimental results of the transient conditions with a good agreement. Through this study, it is expected that the MARS-KS code validation for the CANDU safety analysis could be improved significantly. In addition, this study is expected to contribute to developing practical assessment problems and providing a meaningful example for direction to code improvement.

Optimal Main Heat Transport system configuration for a nuclear power plant

Selection of an optimal Main Heat Transport (MHT) system configuration for a nuclear power plant (NPP) is an activity of high significance, as it influences the safety and operational aspects. This paper brings out the methodology for selection of optimal MHT configuration of NPPs which provides enhanced safety along with ease of operation. The criterion for selection of NPP-MHT configuration have been developed for enhancing safety and ease of operation. As a case study, an innovative MHT configuration has been arrived for a channel type Boiling Water Reactor (BWR) which further improves safety. Some of the multiple failure Design Extension Conditions (DEC) scenarios without SCRAM which lead to severe core damage, for commonly deployed MHT configurations, do not encounter any core heat up with this optimal MHT configuration, for a prolonged duration with Emergency Core Cooling System (ECCS). To demonstrate the effect of MHT system configuration on safety, the Natural Circulation Advanced Boiling Water Reactor (NCABWR) has been considered in this study. Though NCABWR is considered for demonstration purpose, similar methodology can be used for devising or conceptualizing an optimal MHT system configuration for any other NPP. The NCABWR is a four inter-connected loop channel type BWR with natural circulation based passive MHT cooling at all power levels. Following the recent DEC events, more emphasis is laid on very low or no radiological impact in the public domain and no requirement of evacuation in the spirit of “more good than harm”. New NPPs like NCABWR come closer to these requirements. With the mostly passive safety systems, NCABWR can avoid and mitigate DEC. The MHT system configuration for such NPPs need to be selected optimally based on the safety enhancement. In this study, optimal NCABWR MHT configuration has been further evolved based on all the Postulated Initiating Events (PIEs) including multiple failures, such as Anticipated Transients without Scram (ATWS), 200% inlet header break Loss of Coolant Accident (LOCA) without reactor trip etc. Several MHT configurations were conceptualized based on comprehensive overall design, layout requirements and reactor specific PIEs like a LOCA with a postulated break in the four partitioned ECCS header compartment. LOCA in one of the four ECCS header compartment required a mandatory interconnection among all the four MHT loops through the Common Reactor Inlet Header (CRIH) without any partition. This analysis and related issues signaled towards requirement of development of an optimal MHT configuration. The objective of the present study is to configure the MHT components and the ECCS in such an optimal formation so as to provide an enhanced performance in terms of heat removal during adverse conditions. This paper brings out clearly the important aspects needed to be considered while selecting the MHT configuration for the NPPs with enhanced safety even for ATWS-DEC scenarios, without compromising the ease of operation. The development of advanced accident tolerant fuels with improved performance, an improved/optimal MHT NPP configuration is inevitable to ensure safety.

Coupling of reactor transient simulations via the SALOME platform

Canadian Nuclear Laboratories has developed a toolset of coupled reactor transient simulation codes for pressure tube heavy water reactors. This paper presents the first demonstration of the toolset’s capabilities, using a hypothetical 20% reactor inlet header break (critical break) scenario. This scenario induces strong feedback between thermalhydraulic and reactor physics phenomena, owing to a positive coolant voiding coefficient. The scenario was simulated using the advanced thermalhydraulics code ARIANT and the reactor physics code NESTLE-C, coupled together using the SALOME platform. This paper summarizes recent advances in ARIANT, the coupling methodology, and the results of the transient simulation.

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

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

An open calculation of RD-14M small-break LOCA experiments using CATHENA code

The open calculations of two Small Break Loss of Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility are performed as an International Atomic Energy Agency (IAEA) activity of International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the thermal–hydraulic analysis code against qualified data for Heavy Water Reactor (HWR) systems. This ICSP started with the first meeting of the participants in Vienna in 2007 November. Two tests were selected for this activity, test B9006, a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection, performed in 1990, and test B9802, a 3-mm inlet header break experiment, performed in 1998, to provide data with full channel power to study boiling in the channels and condensation in the steam generators. The previously performed blind simulations demonstrated that the CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. However, several significant discrepancies between the code predictions and the measurement data were noted, and attributed to the input errors and the code models relevant to post-dryout (PDO) heat transfer phenomena in the heated channels of test B9802. Therefore, the open calculations are performed to allow correction of previous blind calculations in the present study. These open calculations include the Senaratne and Leung model with the optimized coefficients of its correlation for the test B9802, which can predict more accurately the onset of sustained dryout on the surface of the heater rods. The improvement of the open calculation results are shown by comparing them with the test data and also with the blind calculation results.

A blind simulation of RD-14M small-break LOCA experiments using CATHENA code

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down. These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.

Uncertainty analysis of LBLOCA for Advanced Heavy Water Reactor

The main objective of safety analysis is to demonstrate in a robust way that all safety requirements are met, i.e. sufficient margins exist between real values of important parameters and their threshold values at which damage of the barriers against release of radioactivity would occur. As stated in the IAEA Safety Requirements for Design of NPPs “a safety analysis of the plant design shall be conducted in which methods of both deterministic and probabilistic analysis shall be applied”. It is required that “the computer programs, analytical methods and plant models used in the safety analysis shall be verified and validated, and adequate consideration shall be given to uncertainties”. Uncertainties are present in calculations due to the computer codes, initial and boundary conditions, plant state, fuel parameters, scaling and numerical solution algorithm. All conservative approaches, still widely used, were introduced to cover uncertainties due to limited capability for modelling and understanding of physical phenomena at the early stages of safety analysis. The results obtained by this approach are quite unrealistic and the level of conservatism is not fully known. Another approach is the use of Best Estimate (BE) codes with realistic initial and boundary conditions. If this approach is selected, it should be based on statistically combined uncertainties for plant initial and boundary conditions, assumptions and code models. The current trends are going into direction of the best estimate code with some conservative assumptions of the system with realistic input data with uncertainty analysis. The BE analysis with evaluation of uncertainties offers, in addition, a way to quantify the existing plant safety margins. Its broader use in the future is therefore envisaged, even though it is not always feasible because of the difficulty of quantifying code uncertainties with sufficiently narrow range for every phenomenon and for each accident sequence. In this paper, uncertainty analysis for the Large Break LOCA (200% Inlet Header Break) of Advanced Heavy Water Reactor (AHWR) has been carried out. The uncertainty analysis was carried out for the peak cladding temperature (PCT), based on the two different methods i.e., Wilk’s method and the response surface technique. Their findings have also been compared.

Enhancement of Safety Analysis Capability for a CANDU-6 Reactor Using RELAP-CANDU/SCAN Coupled Code System

The purpose of this study is the development and verification of the coupled code system SCAN and RELAP-CANDU for transient analysis of a Canada deuterium uranium (CANDU) reactor. For this purpose, a spatial kinetics calculation module is developed and implemented in SCAN, a three-dimensional (3-D) CANDU–pressurized heavy water reactor neutronics design and analysis code. Then, a dynamic linked library of the SCAN code is generated for the integration with RELAP-CANDU.The RELAP-CANDU code has been developed for best-estimate transient simulation of CANDU reactor coolant systems based on the RELAP5 code. The SCAN code is a 3-D neutronic calculation code, which is composed of both unified nodal methods based on coarse-mesh finite difference method solutions to the time-dependent two-group diffusion equations.To verify the reliability of the coupled code system RELAP-CANDU/SCAN, the 40% reactor inlet header break accident, the 100% reactor outlet header break accident, and the pump suction pipe break are analyzed. The proposed coupled thermal-hydraulic and neutronic analyses methodology shows that there is an important margin in the traditional accident analysis.

Progress of the DUPIC Fuel Compatibility Analysis—III: Reactor Safety

The compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactor (DUPIC) fuel with the existing 713-MW(electric) CANDU (CANDU-6) reactor has been analyzed for large-break loss-of-coolant-accident (LOCA) scenarios such as a 55% pump suction break, a 35% reactor inlet header break, and a 100% reactor outlet header break. The compatibility was assessed for the fuel integrity against the stored energy and the radiation environmental effect resulting from the large-break LOCA. The analysis showed that the stored energy of the DUPIC fuel was below the fuel breakup energy by 32%. The environmental effect was estimated for the personal and public doses using the radiation source term obtained from one-fourth of the fission product inventory in the fuel gap of the CANDU-6 reactor, being steadily operated at full power. The analyses have shown that both the personal and population doses are below the design limits even for a postulated dual failure such as a complete loss of containment building isolation logic.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Three-dimensional numerical calculations have been performed for a transient moderator circulation inside the CANDU (Canada Deuterium Uranium) calandria vessel of Wolsong Units 2/3/4. The porous media approach was applied for the core region containing 380 calandria tubes. An anisotropic hydraulic resistance model for the porous media has been developed based on the empirical pressure loss correlations. The selected event was the 35% RIH (Reactor Inlet Header) break with a loss of ECC (Emergency Core Cooling) injection, which has been known to give the largest heat load to the moderator among all the DBA's (Design Basis Accidents). The calculation has been successfully done until 1,200 s after the break, when most of the considerable heat transfer procedure has been completed. During this LOCA (Loss of Coolant Accident) transient, the local subcoolings in the vicinity of any PT/CT (Pressure Tube/Calandria Tube) contact does not drop below the experimentally derived subcooling threshold of 30°C. Because the minimum subcoolings reach only a few degrees to the threshold temperature during the initial 20-40 s, future work on the CANDU moderator circulation needs to be aimed at determining whether this small subcooling margin covers the uncertainty of the moderator analysis.

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

Moderator Analysis of Wolsong Units 2/3/4 for the 35% Reactor Inlet Header Break with a Loss of Emergency Core Cooling Injection

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test.

Fuel Channel Analysis for a Large-Break Loss-of-Coolant Accident in a Canada Deuterium Uranium Reactor Loaded with CANFLEX Fuel Bundles

The CATHENA ``slave`` channel model is used for fuel channel analysis of a 30% reactor inlet header break in a Canada deuterium uranium (CANDU)-6 reactor loaded with 43-element bundles of advanced CANDU [CANDU flexible fueling (CANFLEX)] fuel. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The maximum fuel centerline and sheath temperatures for the CANFLEX bundle are lower by 388 and 128C, respectively, than those for the standard bundle because of the lower maximum linear power of the CANFLEX bundle. The pressure tube (PT)/calandria tube (CT) contact for the CANFLEX bundle occurs 2 s later than that for the standard bundle. The PT/CT contact temperature for the CANFLEX bundle is 7C lower than that for the standard bundle. These provide the CANFLEX bundle with a slightly enhanced safety margin for fuel channel integrity in the CANDU-6 reactor, compared with the standard bundle. The effect of bearing pad (BP)/PT contact on the PT temperature predictions is assessed. A BP/PT contact conductance of 3 kW/m{sup 2} {center_dot} K after PT ballooning does not create any hot spot because it gives the contacted PT sector approximately the same heat transfer as convective heating by the hotmore » coolant for the adjacent sector. The assumed BP/PT contact conductance does not threaten the fuel channel integrity.« less