Advanced Passive Pressurized Water Reactor Research Articles

Technical Evaluation Method of Mechanical Equipment Item Substitution in the Procurement and Construction Process of Nuclear Power Plant

Abstract The needs of item substitution are frequent during the procurement and construction of nuclear power project. But the documents prepared for item substitution requests vary considerably in quality that influence working efficiency and project progress. Analyses have been performed for various item substitutions in accordance with laws, regulations, standards, and engineering practices. Focusing on mechanical equipment item substitution, the relative evaluation methods during procurement and construction have been developed based on lessons learned from practices during the engineering practices of advanced passive pressurized water reactor (APPWR) built in China. The methodology of item substitution evaluation can be summarized into six steps, which are critical characteristics for design, nonspecific/specific item substitution, analysis of item function(s) and failure mode(s), impact on related structures, systems and components (SSCs), compensation and coordination measure, and engineering feasibility analysis, respectively. For nuclear power units having been completed, typical cases of item substitution are suggested to be accumulated, and lists of typical critical characteristics for design for the items that often need to be substituted are suggested to be summarized. And then an application library including cases and lists will be established, which will guide item substitution work high quality and efficiency.

Study on properties of 18MND5 steel forgings for PWR steam generator

In the pressurized water reactor (PWR) nuclear power plant, the shell material of steam generator is required to have good strength-toughness matching, anti-fatigue performance, and neutron radiation resistance to ensure the long-term safe and reliable service. The manufacturing requirements of Manganese-Nickel-Molybdenum alloy steel forgings of steam generator channel head for Hua-long Pressurized Reactor (HPR1000) and Advanced Passive PWR(AP1000) nuclear power plant were compared. The heat treatment process, chemical composition, mechanical properties, metallographic structure of 18MND5 steel forging and SA 508-III steel forging were analysed and studied. The results show that the performance heat treatment temperature of HPR1000 forging has more stringent controls and with lower element content of C, Mo, P, Si, V, Co than the AP1000 forging. By reducing the C content, HPR1000 forging got better toughness while the strength was ensured.

Aerosol pool scrubbing phenomena during the containment depressurization venting. Part I: Scaling analysis

For the Chinese Advanced Passive Pressurized Water Reactor, one of the possible containment depressurization measures is to discharge to the Spent Fuel Pool (SFP) during severe accidents. During containment depressurization venting process, carrier gas with aerosol is injected into the SFP from the containment through a discharge sparger. To obtain the aerosol retention effect during this process, a scaled-down test facility have been designed and built. In this paper, the Phenomena Identification and Ranking Tables (PIRT) in the discharge pipe and the SFP is developed. Then, based on the Hierarchical Two-tiered Scaling analysis (H2TS) method, the scaling analysis of SFP scrubbing phenomena is completed. A jet flow model and a particle retention model are developed to analyze these important phenomenon, and a series of similarity criteria are obtained. Based on these similarity criteria, the CASSI test facility is designed, and the scaling rationality is evaluated.

Influences of some engineered factors on IVR-ERVC limits

In-vessel retention (IVR) through passive external reactor vessel cooling (ERVC) has been adopted as one of the key severe accident mitigation strategies for Chinese-designed large-scale advanced passive pressurized water reactor. From the view of thermal load effectiveness, critical heat flux (CHF) of reactor pressure vessel (RPV) out wall has been taken as one of the key criteria for IVR-ERVC feasibility. In order to investigate the ERVC limits, as well as its performance during IVR-ERVC process, a campaign of CHF tests under real plant conditions is conducted on the full-height REPEC-II facility, which has been designed according to the prototypic ERVC configuration and verified following an integrated, hierarchical scaling approach. Based on the abundant results acquired in the tests, it is attempted, in this paper, to summarize and evaluate ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influences on CHF of RPV out wall with various non-uniform heat load distribution and ERVC channel configurations, effects of subcooling, heating surface conditions (prototypic surface material and its oxidization) and water chemistry on external cooling limits, etc. These are expected to help designers with reliable estimation on impact of some related engineered factors on real IVR-ERVC performance.

The Necessity to Reexamine Key Component Material With Spectrometry Analysis in the Nuclear Island

In order to avoid the misuse of metal material in nuclear projects, typical cases happened in advanced passive pressurized water reactor (AP1000) nuclear power projects in China are analyzed. The analysis finding indicates that some cases were caused by defective procedures or undemanding processes performance, and all cases are found to be relevant to human error. It is considered that procedural management cannot completely avoid the misuse of metal material when it is caused by human error, and spectrometry analysis is suggested to reexamine the material of key components.

Full scale experimental study of liquid entrainment at ADS-4 branch of CAP1400

Abstract Liquid entrainment at the tee junction of hot leg to ADS-4 branch (fourth-stage Automatic Depressurization System) is a high priority safety related phenomenon in Westinghouse Advanced Passive pressurized water reactor (AP1000) and Chinese Advanced Passive pressurized water reactor (CAP1400). Previous studies are mainly carried out on reduced scale experimental test facilities. To offer a better understanding of the liquid entrainment phenomenon at tee junction and evaluate the applicability of entrainment models, a full scale test facility is designed and fabricated. The maximum air and water flowrates are 12 kg/s and 150 kg/s respectively. 16 onset of liquid entrainment tests and 80 entrainment rate tests are carried out on this test facility. Test data is compared with previous entrainment models. Significant differences are founded. This means the applicability of liquid entrainment models in safety analysis code is questionable. Based on the flow regime transition criterion in horizontal pipe, a new onset of liquid entrainment model is proposed. The agreement of the new model to test data is better than previous models.

Using Thorium Based Fuel in AP1000: Steady State Analysis

The feasibility of utilizing Thorium-Plutonium Mixed Oxide in the Westinghouse AP1000 Advanced Passive pressurized water reactor is examined under steady-state, beginning of life conditions. Initial core loading of the reactor has three types of UO2 fuel assemblies with the U235 enrichments as: 2.35w/o, 3.40w/o and 4.45w/o. In this paper, one-third of the assemblies are replaced by (Th-Pu)O2 fuel in two arrangements; the first with a blanket of (Th-Pu)O2 fuel replacing the 4.45% enriched UO2 assemblies surrounding the low enriched UO2, and in the second some of the UO2 assemblies are replaced in a way creating a ring of (Th-Pu)O2 fuel in the core. The reactor is modeled using QUARK code. The required cross-section data for QUARK have been generated using WIMSD5. The results of the steady state analysis show that introducing (Th-Pu)O2 fuel into AP1000 would not negatively impact the reactor's safety as the criteria mentioned in design control document are met. For (Th-Pu)O2 blanket and ring loading arrangements, the power peaking factor is less or equal the design limit. Over the length of the hot channel, the MDNBR varies with a minimum above the design limit for blanket and ring loading arrangements. This work provides the basis of the study on fuel behavior and thermal hydraulic analysis of AP1000 to evaluate the safety aspect of various core loading patterns under anticipated and accidental conditions.

Performance verification of Passive Accident Mitigation Scheme (PAMS) under feedwater line break (FLB) accident

An innovative secondary circuit passive residual heat removal scheme, termed as Passive Accident Mitigation Scheme (PAMS), is investigated in our work. PAMS aims at removing core residual heat and ensuring reactor safety for Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP), which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Residual Heat Removal System (PRHRS).Shi et al. (2015) analyzed the PAMS operation characteristics under Station Blackout (SBO) accident using RELAP5. In order to assess the ability of PAMS, more analyses should be carried out. This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and proposes the simulation performance verification of PAMS under FLB accident with the help of RELAP5. Moreover, the comparative analysis is also conducted between PAMS and traditional secondary circuit PRHRS with different design capacities under FLB accident. Results show that: during the whole accident process, the key parameters’ performances of primary system could meet acceptance criteria requirements, which proves PAMS can cope with FLB accident; moreover, since only the PAMS train connected to intact circuit can operate to mitigate FLB accident, the analysis results prove that PAMS design satisfies single failure criterion; the primary system cooling process with PAMS operation is quite close to primary system cooling process with 4% of Full Power (FP) design capacity PRHRS operation, and both of them can remove core residual heat significantly under FLB accident.

Hydrogen risk for advanced PWR under typical severe accidents induced by DVI line break

Advanced passive pressurized water reactor (APWR) relies on In-Vessel Retention (IVR) of molten core debris under severe accidents, which makes hydrogen generation and distribution different with other PWRs. Hydrogen generation and hydrogen risk for APWR is studied with the integrated severe accident plant model, including Reactor Coolant System (RCS), engineered safety features (ESFs), simplified secondary side, containment and passive containment cooling system (PCCS). Initial events of DVI line break with passive core cooling system (PXS) compartment not flooded (named Case 1), DVI line break with PXS compartment flooded through broken DVI line (named Case 2), and DVI line break with IRWST direct injection available based on Case 1 (named Case 3) are selected to study hydrogen generation in pressure vessel, distributions in containment, and hydrogen risk controlled with igniters. Results show that hydrogen generated in pressure vessel is 257kg, 461kg and 499kg for Case 1 to Case 3 due to different thermal hydraulic characteristics. During maintaining the water level in the pressure vessel with the safety measures, zirconium-steam reaction is violent and the duration of hydrogen generation is longer than the other two cases due to the availability of plenty of steam, resulting in an amount of hydrogen generation, and hydrogen distribution in upper containment is nearly 10%. Case 3 is screened to analyze the effectiveness of hydrogen control system consisting of igniters, showing that the hydrogen control system can mitigate hydrogen risk when igniters are available at the SAMG inlet condition.

Study on Heat Removal Capacity of PCCS of Advanced Passive PWR during Severe Accidents

In the advanced passive pressurized water reactor, a passive containment cooling system (PCCS) has been adopted to cool the containment—comprising a cylindrical steel vessel—during postulated accidents, whereby the decay heat is removed through water film evaporating enhanced by air cooling outside the containment. In this study, an integrated safety analytical code is used to study the heat removal capacity of PCCS during severe accidents and its influence on severe accident management measures. The coupled analytical model includes the reactor cooling system, engineered safety features, containment system, and PCCS. Containment responses during typical design-basis accidents and integrated severe accident scenarios are calculated and validated using a design control document and probabilistic risk assessment, respectively. Four typical severe accident sequences that contribute to core damage frequency or containment high pressure are selected to evaluate the containment response. The results show that the containment pressure can be controlled at a relatively low level within 72 h with the heat removal by PCCS. Analysis of the effects of PCCS water cooling recovery during the late period of the accident sequence in severe accident management guidelines alerts as to the risk of hydrogen combustion after breaking the steam-inert atmosphere inside containment. Moreover, sensitivity analysis has been performed to study the influence of the water film coverage rate and environmental air temperature, and it shows that a decrease of the water film coverage rate and an increase of the environmental air temperature reduce the PCCS cooling capacity.

Evaluation of RCS injection strategy by normal residual heat removal system in severe accident management

Severe Accident Management Guidelines (SAMGs) suggests mitigating the consequence of severe accident scenarios by using the non-safety systems if the safety systems are unavailable. For 1000MWe advanced passive pressurized water reactor (PWR), the normal residual heat removal system (RNS) is proposed to implement the Reactor Coolant System (RCS) injection strategy during severe accidents if safety systems fail. Therefore, evaluation of the effectiveness and negative impact of RNS injection strategy is performed, in which two typical severe accident sequences are selected, which are the typical low-pressure core melt accident sequence induced by Large-break Loss of Coolant Accident (LLOCA) with double-ended guillotine break at cold leg and the typical high-pressure core melt accident induced by Loss of Feed Water (LOFW), to analyze RCS response using the integrated severe accident analysis code. The plant model, including RCS, Engineering Safety Features (ESF), containment and RNS, is built to evaluate the effectiveness of RNS injection by comparing the sequences with and without RCS injection, which shows that RNS injection can terminate core melt progression and maintain core cooling in these accident sequences. However, hydrogen generated during the core reflooding is investigated for the negative impact, which shows that RNS may increase the hydrogen concentration in the containment. For the sequence induced by LOFW, two different RCS depressurization measurements are compared, which shows that opening ADS stages 1–3 valves can cause a rapid increase of hydrogen concentration in the In-containment Refueling Water Storage Tank (IRWST) compartment and may lead to hydrogen detonation risk if hydrogen igniters are unavailable, while hydrogen can be well dispersed into the containment by intentionally opening ADS stage 4 valves, which is suggested as the preferred measure.

Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no-load conditions

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Simulation of fission products behavior in severe accidents for advanced passive PWR

Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences.

Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

Evaluation of containment venting strategy via VFS path for advanced passive PWR NPP

Containment would be suffered by over-pressurization if passive containment cooling system (PCS) is failed to remove the heat for the advanced passive pressurized water reactor (PWR). Containment depressurization should be evaluated. Severe accident sequence induced by large break loss of coolant accident (LB-LOCA) at belt of vessel is selected to evaluate the effectiveness of containment venting strategy for advanced passive PWR with the integral safety analyses code, focusing on containment behaviors. Different sizes of the venting path and different open/close pressures of venting are investigated to evaluate CsI mass fraction released to the environment through Containment Air Filtration System (VFS). The results show that depressurization strategy by using the path of VFS can mitigate the high pressure risks and prevent the containment failure when PCS can't be initiated. High pressure at start of venting would reduce the operation times for operator to open the VFS path, reduce the radioactive released to the environment and delay the release time. The Containment Filtered Venting System (CFVS) used in generation II PWR can maintain the containment pressure below the design pressure and reduce the radioactive release for the advanced passive PWR. However much more steam is generated by In-Vessel Retention (IVR) measure, the venting capability is not enough to make the pressure at low value and the venting path should be enlarged.