Decay Heat Removal Analysis

The functions of the engineered passive safety systems of the MRX which are significantly simplified are evaluated with the safety analyses. The LOCA analysis as an example presented in the previous chapter shows that core flooding is kept and decay heat removal is performed successively. Corresponding to the Japanese governmental PWR licensing safety review guideline , the analyses of the accidents and the anticipated transient events are conducted. The... 

In the 4S, the decay heat is removed by two systems consisting of the decay heat removal coil installed in the reactor (PRACS) and the natural air ventilation from outside the guard vessel (RVACS). The analysis considers the destruction of PRACS and the RVACS cooling stack by a large falling aircraft. In addition to this extreme severe condition, 50% of the cross sectional area of the RVACS stack is assumed to be blocked. 

Safety evaluation studies have been conducted for confuming the physical phenomena and integrity of the fuel subassemblies, the core internal structures and the heat transport systems during the normal operation, scram transients and the early stage of postulated accidents. On this account, thermohydraulic experiments related to the decay heat removal by natural circulation have been carried out, and the development and validation of the thermohydraulic safety analysis codes is also in progress. 

Structures, systems and components important to safety shonld not be shared between two or more nuclear power reactors. However, if this is done, it should be demonstrated by test, experiments or engineering analysis that aU safety requirements can be met for all reactors in all states. In the event of accident conditions involving one of the reactors, an orderly shutdown and decay heat removal of the other reactors should be achievable. Special consideration should be given to external events which could cause accidents in more than one plant. Common support systems should be able to cope with all the affected reactors. 

The effects of flooding on a nuclear power plant site may have a major bearing on the safety of the plant and may lead to a postulated initiating event (PIE) that is to be included in the plant safety analysis. The presence of water in many areas of the plant may be a common cause of failure for safety related systems, such as the emergency power supply systems or the electric switchyard, with the associated possibility of losing the external connection to the electrical power grid, the decay heat removal system and other vital systems. Details are provided in Ref [6]. 

The previous analysis of passive decay heat removal had been based conservatively (for safety reasons) on theoretically-determined heat losses assuming ideal manufacturing and installation of the insulation and neglecting any penetrations, disturbances and other irregularities. It predicted a maximum temperature of 740 C. Later analyses utilizing the actual heat losses gave an upper temperature of 570°C after about 20 hours. 

The initial conditions of the experiments were the same as in the case of the emergency core cooling system tests described above. The M-pumps were switched off The dampers were opened partially to avoid sodium temperatures falling below 200°C in the cold legs. The heat removed by natural convection at the air side and at the Na-side amounted to 1.1 MW. This value was consistent with predictions of the given boundary conditions. The dampers of the air coolers, opened by simple manual operation, were introduced as an additional heat sink into the analysis of passive decay heat removal. A maximum temperature of 530°C was then reached after about 10 hours. The demonstrated temperatures lay very close to the normal operating temperatures (Figure 3.2). 

Jubault, M., et. al.. Fast Neutron Reactor Safety Reliability Analysis of PHENIX Decay Heat Removal Function, Proc. of ENS Conference on Fast Reactor Safety, Seattle, USA, 1979. 

ABSTRACT Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RET) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP. 

The AHTR uses passive decay-heat-cooling systems. For the analysis herein, an air-cooled passive decay-heat-removal system was examined that is similar to that developed for the General Electric sodium-cooled S-PRISM. The reactor and decay heat cooling system are located in an underground silo. In this pool reactor, decay heat is... 

Transient simulations of SEALER have been carried out using the SAS4A/ SASSYS-1 codes as well as BELLA, a code written specifically for the purpose of safety-informed design of LFRs. Analysis shows SEALER to withstand unprotected withdrawal of a single control rod, loss of forced flow and loss of heat sink, thanks to its low power density, the capability of natmal convection for decay heat removal, and reliance on thermal radiation from the vessel as the ultimate heat sink. 

As described in the previous chapter, the MRX copes with anomaly including accidents by help of the engineered passive safety system The core is flooded by the water-filled containment and the decay heat is removed by the EDRS and the CWCS. To view the function and the transient behavior, the LOG A analysis using RELAP5/mod2[ll] and COBRA-IV[12] codes is presented here as followings. 

Although the PRACS can remove the decay heat under a natural convection mode, it was assumed to be out of work and only the RVACS was available — this conservative analysis was performed to evaluate the heat removal capability of the RVACS. The movable reflector was assumed to be moved down in this event. 

Results of the analysis for the postulated feedwater line rupture show that the capacity of the PRHR heat exchanger is adequate to remove decay heat, to prevent overpressurising the reactor coolant system, and to maintain the core cooling capability. Radioactivity doses from mptures of the postulated feedwater lines are less than those presented for the postulated main steam line break and meet US regulatory criteria. Fiuther consequence analysis will be conducted to demonstrate that the results meet the relevant UK criteria. 

The starting point for the intact circuit assessment is immediately following the ehangeover from steam generator cooling, when the decay heat burden on the normal residual heat removal system is at its maximum. The analysis assumes the worst case availability of safety measures permitted by the Technical Specifications for any intact circuit shutdown mode (see Chapter 16 of Reference 5.6) ... 

Analysis of the heat removal capability of the passive contaimnent cooling system show that the system is able to remove decay heat following Design Basis events and keep contaimnent pressure and temperature below the design values. The analytical models have been vahdated, ineluding through use of testing performed specifically for this purpose (see Section 6.2.1 of Reference 6.1)... 

A maintenance cooling system (MCS) is provided (see Fig. 5.20) for the purpose of removing the decay heat during the maintenance period. Analysis has indicated that, after 20 hours following the reactor shutdown, the MCS can successfully remove generated decay heat. 

The above analysis is based on the assumption that the reactor has been shut down as a result of the depressurization, so that only the decay heat has to be removed. If the shutdown system were to fail in an AGR, the negative temperature coefficient of the reactor would not be large enough to prevent melting of some of the stainless steel fuel cladding. The loss of the neutron absorption associated with the steel could initiate a reactivity transient which would result in a large-scale core melt. This event, however, is extremely unlikely on account of the extensive protective instrumentation and reliable shutdown system of the AGR. 

Analysis shows that the reactor cavity walls can remove enough decay heat in the first 200 hours to ensure both the fuel and RPV do not exceed design limits regardless of the RCCS performance. 

Three trains of the AFS are provided for the backup of these RCPs. It should be noted that the motor-driven RCPs are not credited in the safety analyses just as they are not credited in BWRs. The capacity of a single train is 4% of the rated flow this is determined on the basis of removing the decay heat up to 6% of the rated power by two trains considering a single failure. The AFS also plays the role of reactor core isolation cooling (RCIC) because the main steam is extracted upstream from the main steam isolation valves (MSIVs). The start time of the AFS is determined by reference to the turbine-driven RCIC of ABWRs. The start-up curve to be used in safety analysis is shown in Fig. 6.5 [2]. The influence of its start time on the reactor safety is investigated in Sect. 6.7. 

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