VVER-440 Containment Thermal Hydraulic Analyses With MELCOR and CONTAIN Codes

The Savannah River Site (SRS) Safety Analysis Report (SAR) covering K reactor operation assesses a spectrum of design basis accidents. The assessment includes estimation of the dose consequences from the analyzed accidents. This report discusses the methodology used to perform the dose analysis reported in the SAR and also includes the quantified doses. Doses resulting from postulated design basis reactor accidents in Chapter 15 of the SAR are discussed, as well as an accident in which three percent of the fuel melts. Doses are reported for both atmospheric and aqueous releases. The methodology used to calculate doses from these accidents as reported in the SAR is consistent with NRC guidelines and industry standards. The doses from the design basis accidents for the SRS reactors are below the limits set for commercial reactors by the NRC and also meet industry criteria. A summary of doses for various postulated accidents is provided.

Design and development of a comprehensive program for the assessment and analysis of environmental effects due to the release of radioactive materials from the stack of nuclear installations: A case study in Bushehr nuclear power plant

Nuclear energy has the potential to provide a large amount of green and clean energy throughout the world. However, in the process of nuclear energy development, safety has always aroused a lot of attention, especially after the Fukushima incident, safety has become an important condition restricting the development of nuclear power. In the design process of the third-generation nuclear power plant, the incorporation of passive and active safety concepts has further enhanced the defense-in-depth capabilities and demonstrate the effective safety function in both design base accident and design extension condition. To verify the effectiveness of safety system independently designed for next generation of HPR1000, this paper selected important thermo-hydraulic phenomena from phenomenon identification and ranking table in small break LOCA, designed and built an advanced integrated test facility with height ratio of 1:3, volume ratio of 1:90 and power ratio of 1:52 through proportional analysis method of H2TS. It has the test ability to simulate the natural circulation stage, the pressure relief blowout stage, the injection stage of the refueling water storage tank, and the long-term sump recirculation stage. The test facility adopts full pressure design, with consideration for medium-pressure operation capacity, has fully covered the main loop, safety system, auxiliary system and other important systems of the reactor plant prototype. At the same time, it has the function of coupling operation with the containment vessel test device, which can release the mass energy to the vessel and accept the steam reflux condensation. In terms of isothermal, isobaric and equal physical properties simulation, the test parameters have reached the international advanced level, and can carry out multi-condition tests in design basis conditions and design extension conditions, as well as multi-plant tests in active and passive plant. In addition, the test facility has the ability to provide test data under accident conditions, verify safety analysis procedures, and study the impact of key equipment and main operating parameters on the defense-in-depth function under passive and active systems. The test facility could operate stably under high pressure medium pressure and low pressure. Also, the system is sensitive to small break accidents. The distortion evaluation shows that the parameters of the test facility face a good match with the prototype, which can be used to study the key phenomena and safety characteristics of the safety system and confirm the effectiveness of the safety design of AHPR1000.

Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

The commercial nuclear power industry has developed and applied improved methods for safety and performance management since the accident at Three Mile Island (TMI) in 1979. These include methods for risk management, identification and application of lessons learned, risk informed regulation, and safety culture improvement. Methods have also been developed to identify strategies and procedures to manage severe accidents - e.g. events outside the design basis that formed the envelope for the initial operating license. The combined implementation of all these technical, organizational, and regulatory changes has led to a significant industry-wide improvement in the performance of nuclear power plants in the US since TMI. This paper summarizes these developments in the nuclear industry, describes a recent application to risk informed safety culture assessment for a Canadian nuclear power station, and explores the potential to apply these methods for design, operation, and regulation of offshore facilities as part of the industry response to the Deepwater Horizon accident. Nuclear industry practices prior to Three Mile Island From the beginnings of the commercial nuclear power industry in the United States in the 1960's the primary unifying concept for demonstrating safety was the Design Basis Accident (DBA). A design basis accident is a postulated event that the plant must withstand. The Safety Analysis Report (SAR) for each facility was required to demonstrate that the plant could withstand the occurrence of specific prescribed DBAs. Examples include loss of coolant accidents (LOCAs), reactivity accidents, steam generator tube ruptures, loss of offsite power, etc. In addition to forming the basis for demonstrating that the plant could be operated safely within a prescribed " safe operating envelope,?? DBAs also established (perhaps inadvertently) the basic paradigm for the development of emergency operating procedures. Similarly, the design of instrumentation was intended to provide information up to but not beyond the conditions expected during a Design Basis Accident. Hidden within the emergency procedures were the assumptions that plant operators would be able to accurately diagnose the event in progress, and their major role would be to monitor the performance of automatic systems and only intervene when automatic systems failed to actuate or to restore normal conditions once the automatic systems had carried out their assigned functions. Finally, there was likely an unconscious assumption that events more serious than the design basis accident would not (or perhaps could not) occur, and that if a plant could withstand the DBAs then safety was assured for other conceivable accident sequences. Unfortunately, as we shall see later, these numerous, often unspoken assumptions were fundamentally flawed.

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

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis

The Anticipated Transient Without Scram (ATWS) is classified as a Category II anticipated transient event in nuclear power plants and represents an important subject in reactor safety analysis. Due to the highly integrated configuration of small modular reactors (SMRs), their system responses under accident conditions differ significantly from those of conventional land-based pressurized water reactors. In this study, the thermal-hydraulic system analysis code Relap5 is applied to establish a nodalized model of a representative SMR. On the basis of verified steady-state conditions, a series of transient simulations under design basis accidents (DBAs) combined with ATWS scenarios are performed to evaluate the inherent safety features of the reactor and the accident mitigation capability of its passive safety systems. The results indicate that, for a reactivity insertion accident (RI), the reactor core exhibits strong self-regulating behavior, with power rapidly returning to the steady-state level. In the case of a loss-of-coolant flow accident (LOCF), the large negative feedback coefficients enable the core to reach hot shutdown within a short period, and the decay heat is effectively removed through natural circulation and the passive residual heat removal heat exchangers (PRHRs). For a small break loss of coolant accident (SBLOCA), the passive injection system continuously injects coolant into the core by gravity after the riser section is emptied, ensuring that the upper part of the fuel remains submerged throughout the transient.

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

Providing a reliable upper limit of radiological consequences to the plant personnel and the general public is typically the aim of a safety evaluation for anticipated operational occurrences or design basis accidents, as presented in a safety analysis report. A typical tool for dispersion calculation and dose evaluation is MACCS2. In the present analysis four types of calculations are presented: a first calculation, typical for licensing analysis, with the MACCS2 computer code. In a second step conservative assumptions e.g. ground release even if a stack release would be realistic, are dropped. In a third step calculation two is repeated with RODOS, a code (online decision making tool) used to predict the radiological consequences of an accidental release of activity. The step three calculation still contains all the conservative assumptions that are built in the MACCS2 code. In a last step these assumptions are removed, and a “best estimate” calculation on the dose to the public is performed. The whole analysis (step one to four) is repeated for different source terms (noble gases only, tritium dominated, primary system water …) and for different weather conditions. Two main conclusions can be drawn. The first by comparing step two (MACCS2) and step three (RODOS). Here the boundary conditions of the calculations are set to be as similar to each other as possible. The paper shows that despite the fact that MACCS2 uses a Gaussian plume model, while RODOS uses a puff model for dispersion calculation, doses of the same order of magnitude are calculated. For the second conclusion the step one (MACCS2, conservative) and step four (RODOS, best estimate) calculations are compared, it is shown that although the margin of conservatism varies considerably from case to case, the results differ at least one order of magnitude.

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR’s surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip.

The Integral Fast Reactor (IFR) program requires irradiation of a number of U-Pu-Zr metallic alloy elements, to obtain experimental information prior to large scale production of this fuel. The IFR Experimental Fuels Laboratory (EFL) will be established for the fabrication of a limited number uranium-plutonium-zirconium (U-Pu-Zr) alloy fuel elements, as well as the development of process concepts and parameters needed to make the fuel. The EFL will be located in the Analytical Laboratory, which is a low-hazard facility as determined by its Safety Analysis Report (SAR). The U-Pu-Zr fuel will be prototypical of the fuel that will be used in the Integral Fast Reactor (IFR) Program. Approximately 150 U-Pu-Zr alloy pins will be fabricated within the facility. The pins will be processed, inspected, and finally placed into element jackets that have been preloaded with sodium in another facility. After settling, the elements will be transferred to the Fuels and Subassembly Storage Building (FASB) for closure welding and process completion. The purpose of this document is to evaluate the effects to the public and/or working personnel of any incident in the EFL, which would release radioactive contamination to the environment. Several types of probable incidents that could occur within the EFL will be addressed along with actions that will be taken to prevent their occurrence. The document will conclude with an analysis of the most probable design basis accident (DBA), its radiological impact, and also a short discussion of a proposed maximum hypothetical accident. (MHA).

To assess the structural integrity of reactor vessels (RVs) during pressurized thermal shock (PTS) events, a deterministic fracture mechanics (DFM) approach has been widely used such as the procedure in JEAC4206-2007. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RV has become attractive recently because uncertainties related to input parameters can be incorporated rationally. The probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. In this paper, in order to verify the applicability of a PFM method to JEAC4206-2007, deterministic and probabilistic analyses have been performed, and the effects of initial crack size defined in JEAC4206-2007 on the temperature margin obtained from DFM and the probability of crack initiation obtained from PFM have been evaluated. With regard to the PTS event variation, a stuck open valve scenario (SO) has been considered in addition to large- and small-break loss of coolant accident (LBLOCA, SBLOCA) and main steam line break (MSLB).

Small break loss of coolant accidents: Bottom and side break