Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.     

A minimum hot excess core design strategy for ESBWR

The Economic Simplified Boiling Water Reactor (ESBWR) is GEH’s next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. As part of design simplification it uses natural circulation flow with no recirculation pumps or their associated piping. The control blades are the primary control mechanism to address the need for performing reactivity adjustments (using fine-motion drives) at or near rated steady state power. This introduces the potential for duty-related fuel failure, which has to be rigorously addressed as part of reliable design and operation. As means to mitigate this potential for duty-related fuel failure and also to support a simplified ESBWR operation, this study investigates the feasibility of a fuel cycle core design strategy. The objective is to design fuel bundles, and to use them for developing a core design, that minimizes (but does not eliminate) the use of control blades during operation. The reduction in use is envisioned in their number as well as movement in the core. In such a strategy, the effect of the burnable poison in the fuel (that largely drives the core reactivity) is enhanced, and operationally the control blades react modestly to maintain the core critical. While the logic is simple, challenges exist in developing such a design because it needs to balance the requirement for having enough blade inventory in the core to address design/operational constraints and uncertainties. The strategy is conceptualized as “minimum hot excess (reactivity)” design. It reduces the number of blades in the core during normal operation by 50% in comparison to a similar fuel cycle core design with regular inventory of control blades. Because of the increased burnable poison, the minimum hot excess core design strategy comes at a cost of fuel cycle efficiency. This cost is determined in terms of an increased enrichment for the fresh fuel batch fraction.     

Passive containment cooling system (PCCS) response with Drywell Gas Recirculation System (DGRS) activated during a postulated Loss of Coolant Accident (LOCA)

Passive Containment Cooling Systems (PCCS) are characterizing the design of several advanced LWR such as SBWR, ESBWR, ABWR-II, etc. These systems should ensure the mitigation of postulated accidents both under Design Basic Accident (DBA) and Beyond DBA (BDBA) conditions. Some ALWR designs integrated in the PCCS a system called Drywell Gas Recirculation System (DGRS). The DGRS works like a fan, with inlet flow lines connected to the Passive Cooling condenser (PCC) vent lines and the outflow line connected to the Drywell (DW). The present paper presents the experimental results of an integral containment test performed in the PANDA facility. The initial conditions (temperature, pressure, gas composition, decay heat, etc.) for the test represent the containment situation 1 h after a Loss of Coolant Accident (LOCA). The test consists of two phases (6 h each) for a total duration of about 12 h. In the first phase has been simulated the response of the PCCS to a LOCA, in the second phase the DGRS has been activated and has been investigated the effect of such activation on the overall PCCS response. The test shows that the activation of the DGRS has an effect on the overall PCCS characteristics, i.e. composition of gas mixture in the PCC tubes, stratification in the Wetwell (WW), DW-WW pressure differences, timing for the opening of the Vacuum Breakers (VB) and overall containment pressure.     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

Primary loop study of a VVER-1000 reactor with special focus on coolant mixing

The intention of this work is to demonstrate the ability of modern computational fluid dynamics (CFD) tools like ANSYS CFX to model technical facilities with complex geometry such as a reactor pressure vessel with its primary loops over all relevant scales from a range of approximately 1 mm up to a largest dimension of about 50 m. For this purpose, a detailed model of a VVER1000 reactor pressure vessel (RPV) and an extended model with simplified primary loops (Loop model) are presented. The primary loop components like steam generators and pumps are modelled by the outer shapes and additional source terms for energy and momentum exchange. The RPV model part is tested without primary loops; results are compared with former results obtained with a coarser model. The improved model shows less sensitivity on the discretization scheme, and recalculated hot leg temperatures transients are improved. As an example case for the Loop model, the swirl of flow patterns at the core inlet derived from experimental data obtained at the Kozloduy Nuclear Power Plant (Unit 6) is simulated. As progress to the RPV model the Loop model is able to predict a swirl, but due to the lack of technical details, like the pump impellers, the swirl is overestimated.     

Characterization of neutron-irradiated ferritic model alloys and a RPV steel from combined APT, SANS, TEM and PAS analyses

Understanding the behavior of reactor pressure vessel (RPV) steels under irradiation is a mandatory task that has to be elucidated in order to be able to operate safely a nuclear power plant or to extend its lifetime. To build up predictive tools, a substantial experimental data base is needed at the nanometre scale to extract quantitative information on neutron-irradiated materials and to validate the theoretical models. To reach this experimental goal, ferritic model alloys and French RPV steel were neutron irradiated in a test reactor at an irradiation flux of 9 × 10¹⁷ nm⁻² s, doses from 0.18 to 1.3 × 10²⁴ nm⁻² and 300 °C. The main goal of this paper is to report the characterization of the radiation-induced microstructural change in the materials by using the state-of-the-art of characterization techniques available in Europe at the nanometre scale. Possibilities, limitations and complementarities of the techniques to each other are highlighted.     

Temperature evaluation of an instrumented capsule after material irradiation tests in HANARO

An instrumented capsule has been used for an irradiation test of various nuclear materials in the research reactor, HANARO. The capsule is designed to have a standard 4-hole structure for the economical test of an RPV material at 290 ± 10 °C. The temperature of the specimens for the reactor powers, 0-24 MW, is measured by 12 thermocouples, and finite element (FE) analyses are also performed to compare and verify the irradiation test results. As a result of the tests and analyses, the maximum temperature at the reactor power of 24 MW is 256 °C for an irradiation test and 202.6 °C for an FE analysis at Stage 3 of the capsule. Also, for each stage of the capsule, the temperature difference of the specimen in the axial direction is very small to within 10 °C. It is expected that the results presented in this paper will be useful when designing the instrumented capsules for an irradiation test.     

Failure behavior of Zircaloy-4 cladding after oxidation and water quench

Simulated LOCA (loss of coolant accident) tests and subsequent mechanical tests on Zircaloy-4 cladding were carried out to evaluate the failure behavior of the cladding. Zircaloy-4 claddings were oxidized in a steam environment from 900 to 1250 °C for a given time period followed by a flooding of cool water to simulate LOCA tests. After the simulated LOCA test, the ductility of the oxidized cladding was evaluated by mechanical tests such as ring compression test and 3-point bend test. Evaluation of the absorbed contents such as hydrogen and oxygen were also carried out. The results showed that Zircaloy-4 cladding failed during thermal shock when the ECR (equivalent cladding reacted) value exceeded 20%. Lower boundary of brittle failure at thermal shock corresponds to 20% of ECR line calculated by the Baker-Just equation regardless of test temperature. On the other hand, boundary of ductile failure by the mechanical test did not followed after the ECR line. It rapidly decreased above 1000 °C to show that all Zircaloy-4 claddings behaved brittle fracture above 1150 °C when it oxidized at 300 s. Microstructural analysis revealed that boundary of ductile failure by the mechanical test fitted well when the absorbed oxygen content inside the prior-β layer was below 0.5 wt%.     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

Severe accident management strategy verification for VVER-1000 (V320) reactor

Severe accident analysis of a reactor is an important aspect in evaluation of source term. This in turn helps in emergency planning and Severe Accident Management (SAM). The use of the Severe Accident Management Guideline (SAMG) is required for accident situation which is not handled adequately through the use of Emergency Operating Procedures (EOP), thus leading to a partial or a total core melt. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). Initiation of SAMG for VVER-1000 is considered at two core exit temperatures viz. 650 °C as a desirable entry temperature and 980 °C as a backup action. Analyses have been carried out for VVER-1000 (V320) for verification of some of the strategies namely water injection in primary and secondary circuit. These strategies are analysed for a high and low pressure primary circuit transients. Station Black Out (SBO) is one such high pressure transient for which core heat can be removed by natural circulation of the primary circuit inventory by maintaining the secondary side inventory. This strategy has been verified where the feed water injection to secondary side of SG is considered from external power sources (e.g. mobile DG sets) as suggested in SAM guidelines. The second transient, a low pressure event is analysed for verification of the SG flooding and core flooding strategies. The analysis shows that SG flooding is not adequate to arrest the degradation of the core. In case of core flooding strategy, the analyses show that core flooding is not adequate to arrest the degradation of the core for the large break LOCA where as for small break LOCA the injections through available safety systems are adequate. The assessments are carried out with integral severe accident computer code ASTEC V1.3.     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

Experimental investigation of natural circulation instability in a BWR-type small modular reactor

The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV). Specifically, the NMR is one third the height and area of a conventional BWR RPV with an electrical output of 50 MWe. Experiments are performed in a well-scaled test facility to investigate the thermal hydraulic flow instabilities during the startup transients for the NMR. The scaling analysis for the design of natural circulation test facility uses a three-level scaling methodology. Scaling criteria are derived from non-dimensional field and constitutive equations. Important thermal hydraulic parameters, e.g. system pressure, inlet coolant flow velocity and local void fraction, are analyzed for slow and fast normal startup transients. Flashing instability and density wave oscillation are the main flow instabilities observed when system pressure is below 0.5 MPa. And the flashing instability and density wave oscillation show different type of oscillations in void fraction profile. Finally, the pressurized startup procedure is recommended and tested in current research to effectively eliminate the flow instabilities during the NMR startup transients.     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

UPTF experiment: Flow phenomena during full-scale loop seal clearing of a PWR

To investigate the flow phenomena in the primary system of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) occurring with a small or intermediate break, experiments were performed at the full-scale Upper Plenum Test Facility (UPTF). Within the Transient and Accident Management (TRAM) program integral and separate effect tests were carried out to study loop seal clearing and to provide data for the further improvement of computer codes concerning the reactor safety analysis. This paper describes the UPTF tests that focus on the sequence of loop seal clearance in a four-loop operation for two different cold leg break sizes and the residual water levels, the flow patterns in, and the pressure drops across a single loop seal during the clearing. The UPTF results obtained from a single-loop seal operation are compared with experimental data and correlations available in the literature. Two correlations are proposed which allow the quantification of residual water levels in the loop seal under PWR conditions. It is shown that the steam–water test results gained from the full-scale UPTF with realistic PWR loop seal geometry differ from those obtained from the full or small-scale test facilities under air–water conditions. The UPTF experiments indicate the substantial need for steam-water test data from a full-scale facility with realistic PWR geometries in order to validate PWR LOCA thermal-hydraulic system codes to predict loop seal clearing correctly.     

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.     

A study on the minimum steam cooling pressure of EOP for BWR power plants

The minimum steam cooling pressure (MSCP) is an important parameter for safe operation of boiling water reactor (BWR)-type nuclear power plant for the anticipated transient without scram (ATWS) scenario with reactor pressure vessel (RPV) water level unknown. Under such situation, the operator is requested to open the safety/relief valves (SRVs) and control the RPV pressure slightly above the MSCP so that adequate core cooling can be maintained. It is derived based on steam cooling strategy.The MSCP, defined to be the lowest RPV pressure at which the covered portion of the core, is capable of generating sufficient steam to preclude peak cladding temperature (PCT) in the uncovered portion of the core from exceeding 1088 K (1500 °F). It is calculated by two parameters - (1) the minimum bundle steam flow (Wg-1500) to maintain PCT < 1088 K (1500 °F) and (2) the number of SRVs available for opening.For current emergency operating procedure (EOP), only one set of MSCP derived based on one value of Wg-1500 for the ATWS condition. Furthermore, it is derived based on decay power of 2.2% rated power. Thus, the current MSCP used for the ATWS accident scenarios was deemed inadequate. The purpose of this paper (work) is to study the MSCP used in the ATWS conditions. For case of ATWS of 13% full power, controlling RPV pressure at MSCP of current approach ends up with core melt. The Wg-1500 is suggested to be replaced by the steam generation rate at minimum steam cooling RPV water level (MSCRWL), which is a function of power level. Simulation result indicates controlling RPV pressure at MSCP is equivalent to controlling the RPV water level at MSCRWL. The revised MSCP is dependent on the ATWS power level.     

Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.     

An evaluation of a general venting strategy in a CANDU 6 reactor building

If the reactor building sprays or local air coolers are not available, depressurization by reactor building venting is considered as a useful mitigation strategy for a severe accident management of the Wolsong plants. As the containment filtered vent system is not established in the Wolsong Units, the reactor building isolation system can be a substitute for reactor building venting. The D₂O vapour recovery system which has a 0.76 m (30 in.) diameter penetration is expected to meet the NRC requirements. To investigate the effectiveness of the Reactor Building Venting Strategy, three kinds of accidents are analyzed: a SBO, a Small LOCA and a Large LOCA. The reactor building pressure behavior was analyzed with the ISAAC computer code for four different cases: without venting, 379 kPa(g)/345 kPa(g) (55 psig/50 psig), 345 kPa(g)/276 kPa(g) (50 psig/40 psig) and 345 kPa(g)/207 kPa(g) (50 psig/30 psig) valve open/close pressures. When the reactor building spray or local air coolers can not be operated, a depressurization strategy by using the D₂O Vapour Recovery System could prevent a reactor building failure and reduce the amount of CsI released to the environment. The present study shows that the operation of valves at a pressure of 379 kPa(g)/345 kPa(g) (55 psig/50 psig) is safe and effective. Based on the current study, the strategy of reactor building venting is involved in severe accident management guidance-5.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.