Auxiliary Feedwater Research Articles

Analysis of PWR SBO sequences with RCP passive thermal shutdown seals

ABSTRACT The operation of recently implanted low-leakage seals after Fukushima has altered the analysis of classical PWR Station Blackout (SBO) sequences , as Seal Loss of Coolant Accident (SLOCA) is no longer one of the dominant factors in the accident progression . An analysis of different management strategies in non-SLOCA sequences has been performed by means of the Integrated Safety Assessment (ISA) methodology using the SCAIS-MAAP model of a 3-Loop PWR Westinghouse design. Through the use of the Damage Domain concept(i.e. the region of the uncertain crew actuation times or physical parameters space where each damage limit is exceeded for each sequence), the times for reaching different damage limits are obtained. Results evidence the positive impact of low-leakage seals, which greatly increase the margin to core uncoveryand reduce core damage frequency. Results also allow concluding that an SBO is dominated, namely by the Auxiliary Feed-Water (AFW) mass flow(turning blind AFW management into an essential procedure), SLOCA (in case the new low-leakage seals fail or they are not present), an excessive AFW mass flow (leading to Turbine-Driven Pump failure) and the DC failure time (losing control valves and the instrumentation).

Integral effect test on station blackout (SBO) scenario with steam generator tube rupture (SGTR) in ATLAS facility

Station blackout (SBO) is one of the most important design extension conditions (DECs) because a total loss of the heat sink leads to a core uncovery or damage without any proper operator action. During a long transient of an SBO, a steam generator tube rupture (SGTR) accident can occur when a steam generator tube is exposed to a superheated steam flow. In this study, a prolonged SBO with an SGTR occurrence was experimentally investigated by performing an integral effect test with the ATLAS facility. For the transient simulation, the tube rupture was simulated when the core water level became below the top of the active core. As an accident management measure, the auxiliary feedwater was supplied to the intact steam generator when the maximum heater surface temperature started to show an excursion behavior. The experimental results showed that a single tube rupture could not sufficiently reduce the primary system pressure before the injection of the auxiliary feedwater. A delayed supply of the auxiliary feedwater after an excursion of the heater surface temperature successfully cooled the primary system until the end of the transient, where inflow of the coolant from the pressurizer contributed to a recovery of the coolant inventory in the core. The natural circulation flow characteristics in the primary system showed an oscillating behavior depending on the heat removal rate of the steam generators. This integral effect test data can be used to evaluate the prediction capability of safety analysis codes and identify code deficiencies in predicting an SBO transient with an SGTR occurrence.

A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

A stochastic hybrid systems model of common-cause failures of degrading components

Common-Cause Failures (CCFs) are an important threat to safety critical systems. Most existing CCF models assume that the component failure behavior does not vary over time. Such an assumption is often challenged in practice due to the influence of various degradation mechanisms, e.g., wear, corrosion, fatigue, etc. In this paper, we develop a new model for CCFs considering components degradation. The model is developed in the mathematical framework of Stochastic Hybrid Systems (SHS). The CCFs are modeled as random shock processes that affect a group of components simultaneously and the components degradation processes are modeled by stochastic differential equations derived from physics-of-failures. The benefit of using the SHS model for CCFs is that the developed model is analytically solvable. The system reliability can, then, also be solved analytically in closed form. The proposed CCF modelling framework is demonstrated by a numerical example of a three-unit redundant system and, then, applied to an Auxiliary Feedwater Pump (AFP) system of a Nuclear Power Plant (NPP). A comparison to the Binomial Failure Rate (BFR) model of literature shows that by considering the components degradation processes, the proposed model can accurately describe the CCF effect on the reliability of a system with degrading components.

Research on capability of secondary passive residual decay heat removal system after Main Feedwater Line Break (MFLB) accident

All advanced NPPs are required to improve safety strategy to cope with design extension condition (DEC) after Fukushima accident. One design improvements for HPR1000 is adoption of secondary Passive Residual Heat Removal System (PRS) to cope with station blackout accident together with loss of all feedwater, which can help to strengthen defense in depth capability and enrich DEC mitigation measures. To make the best use of PRS decay heat removal capability, investigation of PRS startup performance and decay heat removal capability after MFLB accident are investigated based on HPR1000 RELAP5 model. PRS startup response featured by pressure, mass flow rate and void fraction are investigated under three startup strategies. Strategy 3(backflow valve open strategy) is recommended due to minimum fluctuation during initiating stage, in which strategy steam isolation valve is normally open and backflow isolation valve is used to initiate PRS. Then Main Feedwater Line Break (MFLB) accident was simulated with the assumption of loss of auxiliary feedwater. Response characteristics in primary loop and PRS are investigated. After PRS being initiated after scram with 60 s delay, natural circulation in PRS will establish rapidly, and decay heat will be removed under natural circulation operation mode. No overall boiling occurred in primary circuit before isolation of affected SG. Secondary PRS is adequate to remove decay heat after DEC.

Investigation on response of HPR1000 under different mitigation strategies after SGTR accident

Steam Generator Tube Rupture (SGTR) accident was reclassified as Class III Design Basis Accident (DBA) in 1991 due to its higher frequency of occurrence. Under Class III condition, radioactive release safety criteria are even stricter than thermal hydraulic criteria. In this paper, HPR1000 SGTR accident was simulated with two kinds of mitigation measures after steady-state initialization. Key system responses (such as pressure, mass flow rate in primary and secondary system, SG level, short term phase terminating time and radioactive release) were analyzed and compared for the two strategies. No SG overfilled nor other thermal hydraulic limit occurred for both strategies, while lower SG overfill risk and higher radioactive release risk existed for strategy 1 and converse situation existed for strategy 2. Specially, it is recommended VDA to be actuated by SG level, and instant main feedwater and auxiliary feedwater isolations were strongly required for strategy 2. When designing SGTR mitigation measure, there should be a compromised balance between SG overfill risk and radioactive release. Optimized mitigation strategy can be obtained through more accurate activity signal and higher auxiliary feedwater mass flow rate.

The concept of the innovative power reactor

Abstract The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

Simulation of water hammer phenomena using the system code ATHLET

Abstract Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

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.

Evaluation of mechanistic wall condensation models for horizontal heat exchanger in PAFS (Passive Auxiliary Feedwater System)

PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features in the design of the APR+ (Advanced Power Reactor) nuclear power plant, in order to cool down the reactor coolant system without any external supply of electricity or water during an accident. The driving force of the PAFS is condensation inside horizontal tubes of PCHX (Passive Condensation Heat Exchanger). To improve the prediction capability for the condensation heat transfer in the PCHX, advanced wall condensation models which mechanistically consider the wall condensation with the different flow regimes in a horizontal channel were implemented into a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant). The models were evaluated by comparing to the experimental data from the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility, which is a separate effect test to validate the cooling performance of the PCHX. Calculation results showed that the condensation model package (Ahn et al., 2014) presented the enhanced the prediction capability of the SPACE code in predicting the steam pressure and temperature under quasi-steady state conditions. Since this model package is composed of mechanistic models by considering the film condensation in the steam phase and the convection in the condensate liquid separately, it contributed to enhance the prediction capability of the SPACE code and reduce the conservatism in the prediction of the cooling capability of the PAFS. The SPACE code with improvement of the wall condensation model can be applied in predicting the cooling and operation capability of the PAFS more realistically in the safety analysis of APR+ nuclear power plant.

Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding alloys, Cr, ZrSi, TiAlC, and TiSiC. This ATF performance analysis platform will also be used to support experimental work underway in our Integrated Research Project.

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

ABSTRACTDuring a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.

Analysis of SBLOCA on CPR1000 with a passive system

Since the Fukushima accident in 2011, more and more attention has been paid to nuclear reactor safety. A number of evolutionary passive systems have been developed to enhance the inherent safety of reactors. This paper presents a passive safety system applied on CPR1000, which is a traditional generation II+ reactor. The passive components selected are as follows: (1) the reactor makeup tanks (RMTs); (2) the advanced accumulators (A-ACCs); (3) the passive emergency feedwater system (PEFS); (4) the passive depressurization system (PDS); (5) the in-containment refueling water storage tank (IRWST). The model of the coolant system and the passive systems was established by utilizing a system code (RELAP5/MOD3.3). The SBLOCA (small-break loss of coolant) was analyzed to test the passive safety systems. When the SBLOCA occurred, the RMTs were initiated. The water in the RMTs was then injected into the pressure vessel. The RMTs’ low water level triggered the PDS, which depressurized the coolant system drastically. As the pressure of the coolant system decreased, the A-ACCs and the IRWST were put to work to prevent the uncovering of the core. The results show that, after the small-break loss-of-coolant accident, the passive systems can prevent uncovering of the core and guarantee the safety of the plant.

Extending SBO coping capability: An improved auxiliary feedwater system

A novel way to increase the station blackout (SBO) coping capability for pressurized water reactors is presented and evaluated for Advanced Pressurized Reactor 1400 (APR1400). The technical solution proposed is to install an electric direct current (DC) generator coupled to the shaft of a turbine-driven pump (TDP). The new system, named turbine-driven pump – DC (TDP-DC), works as a complement to the plant class 1E batteries, and it is meant to extend the plant coping time beyond present 8-h capacity. Assessment is performed using the one-dimensional multi-dimensional analysis of reactor safety (MARS-KS) system code by evaluating the plant response over an assumption of only one TDP-DC available. The test has been conducted for two conditions: SBOs with/without a reactor coolant pump (RCP) seal loss-of-coolant-accident (LOCA). For both cases, it is assumed to have a mature nuclear core (after three refueling cycles) and a degradation of 20% in efficiency of the only turbine driver working. Results shows that, in 2.7 h from SBO onset, the new TDP-DC system is capable of supporting all electrical loads required during SBO. It is proven that the TDP-DC can extend the plant SBO coping time from 8 to 20 h for the most limiting case of SBO with RCP seal LOCA. In addition, it is revealed that the coping time can be extended beyond 32 h for SBO without RCP seal LOCA.

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

System code improvements for modelling passive safety systems and their validation

Abstract GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구

The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.

Reliability evaluation of auxiliary feedwater system by mapping GO-FLOW models into Bayesian networks

Bayesian network (BN) is a widely used formalism for representing uncertainty in probabilistic systems and it has become a popular tool in reliability engineering. The GO-FLOW method is a success-oriented system analysis technique and capable of evaluating system reliability and risk. To overcome the limitations of GO-FLOW method and add new method for BN model development, this paper presents a novel approach on constructing a BN from GO-FLOW model. GO-FLOW model involves with several discrete time points and some signals change at different time points. But it is a static system at one time point, which can be described with BN. Therefore, the developed BN with the proposed method in this paper is equivalent to GO-FLOW model at one time point. The equivalent BNs of the fourteen basic operators in the GO-FLOW methodology are developed. Then, the existing GO-FLOW models can be mapped into equivalent BNs on basis of the developed BNs of operators. A case of auxiliary feedwater system of a pressurized water reactor is used to illustrate the method. The results demonstrate that the GO-FLOW chart can be successfully mapped into equivalent BNs.

Calculation and updating of Common Cause Failure unavailability by using alpha factor model

The most lethal role of Common Cause Failures (CCFs), which motivate the experts to investigate it, is the dependent behavior therein contained, which leads to simultaneous failure of the systems. Highly redundant systems are more susceptible to be affected by CCFs and also CCFs have been recognized as the principal contributor in the terrestrial reactor accidents. In the past, plenty of work has been done regarding the calculation of unavailability of different types of systems due to CCFs by using different techniques such as fault tree analysis (FTA). But the qualitative aspects such as human errors, maintenance faults and poor components quality cannot be updated by using FTA as the changes occur. So in order to overcome this problem, multinomial distribution function and its conjugate Dirichlet distribution function has been used as likelihood and prior, respectively, in Bayes theorem to obtain an updated posterior function of the same form as Dirichlet distribution function thus improving the working and monitoring capability of Probabilistic Safety Assessment (PSA). Furthermore, the presented research highlights a mathematical model to estimate system unavailability due to CCF by using alpha factor model. By using this model, we can calculate failure probability (unavailability) of the systems quite accurately through the two parameters αk and Qt. The ease of using the proposed model can be assessed through the brief analysis of a case study of Auxiliary Feed Water System (AFWS). AFWS is used in all designs of Pressurized Water Reactor (PWR). It plays a vital role in maintaining a heat sink by providing feedwater to the steam generators.

Modeling of high pressure steam condensation in inclined horizontal tubes of PAFS in APR+

ABSTRACTA Single-tube COndensation exPeriment (SCOP) was performed at Korea Advanced Institute of Science and Technology (KAIST) to simulate the condensation phenomena of Passive Auxiliary Feedwater System (PAFS), which is introduced in Advanced Power Reactor Plus. SCOP tests were conducted for two pipes of different diameters [22.6 and 44.8 mm inside diameter] with 8.2 m length and 2.5–4 degrees of the inclined angles of the upper and lower section of each tube. The experiment were performed at 1.5–6.5 MPa of steam pressure and 0.12–0.45 kg/s of steam flow. Based on the Kim and No correlation (KAIST, 2000), we developed a new turbulent–condensation correlation for the turbulent region of the liquid film in the inclined tube. We introduced the factors to account for the different inclined angle effect, different diameter effect. The Nusselt correlation is adopted for condensation near the tube inlet where the laminar region of the liquid film is dominant due to the small amount of condensed water. The present correlation was compared with the Shah and RELAP5 condensation model using the SCOP data. The root mean square errors of the present correlation, Shah model, and RELAP5 model are 17.41%, 57.88%, and 43.07%, respectively: the present correlation can predict well the condensation characteristics in comparison with other correlations in PAFS.