Feedwater Line Research Articles

Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding temperature effectively. In the MFLB, due to the proper design of the PEFSs, the PRZ was not overfilled by water so that no liquid leakage to the containment happened. Massive radiation release was prevented. Generally speaking, the passive safety system functions as expected, and the three accident mitigation method were proved to be useful for mitigating corresponding accidents.

Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model

In this study, a computational fluid dynamics (CFD) analysis of the transient flow field inside the secondary side of a nuclear reactor steam generator (SG) during blowdown following a feedwater line break (FWLB) accident is performed to evaluate the transient hydraulic loading (pressure) on the SG internals and tubes. The nonflashing liquid flow is assumed for a conservative prediction of the transient blowdown loading. The CFD analysis results are illustrated in terms of the transient velocity and pressure disturbances at some selected monitoring points inside the SG secondary side and compared with those predictions obtained from the existing simple analytical model to examine the physical validity of the CFD analysis model. As a result, the existing simple analytical model cannot yield the transient velocity and pressure disturbances and results in underestimation during blowdown as compared to the CFD calculations. Based on the present CFD analysis results, it is seen that an FWLB may result in excessive disastrous transient hydraulic loading on the SG internal structures and tubes near the feedwater inlet nozzle due to the significant pressure changes (pressure wave with very high amplitude) and abruptly increased velocity of water near the feedwater nozzle.

Estimation of a boundary to distinguish between radioactive materials and non-radioactive materials around a Main Steam line and a Feed Water line in a Biological Shielding Wall of a BWR for decommissioning

In order to estimate amount of radioactive concrete waste from decommissioning activities of a Boiling Water Reactor (BWR), we performed an estimation of a boundary to distinguish between radioactive materials and non-radioactive (NR) materials, which we call NR boundary in our work. So as to estimate the boundary reliably, we measured radioactivities induced by neutrons with activation foils around a Main Steam (MS) line in a MS tunnel room and also calculated three dimensional (3D) distribution of neutron-flux around both the MS line and a Feed Water (FW) line, which penetrate a Biological Shielding Wall (BSW) of the BWR, by using 3D discrete-ordinate method (Sn) calculation code TORT. We estimated the NR boundary by using TORT and verified a validity with comparing the calculation with the measurements. In the calculation, combinations of various numbers of directional mesh sets and orders of scattering cross section were examined by comparing the calculations with the measurements to survey the best set that simulates neutron transport phenomena in the MS tunnel room. The calculation that we judged reliable showed that the NR boundary around these two lines was located in the BSW concrete and that all materials in the MS tunnel room are non-radioactive.

Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

To ensure the operational safety of an industrial water-tube steam boiler it is very important to assess various accident scenarios in real plant working conditions. One of the most challenging scenarios is the loss of feedwater to the steam boiler. In this paper, a simulation of the behavior of an industrial water-tube radiant steam boiler during feedwater line break accident is discussed. The simulation is carried out using the RELAP5 system code. The steam boiler is installed in an Algerian natural gas liquefaction complex. The simulation shows the capabilities of RELAP5 system code in predicting the behavior of the steam boiler at both steady state and transient working conditions. From another side, the behavior of the steam boiler following the accident shows how the control system can successfully mitigate the effects and consequences of such accident and how the evaporator tubes can undergo a severe damage due to an uncontrolled increase of the wall temperature in case of failure of this system.

Cooling performance of PRHR HX in main feedwater line break accident

Cooling performance of PRHR HX in main feedwater line break accident

Behaviors of passive safety systems under a Feed Water Line Break LOCA on a Generation III+ Boiling Water Reactor

A postulated Large Break Loss-Of-Coolant-Accident (LB LOCA) on a generic Generation III+ Boiling Water Reactor (BWR) equipped with passive safety systems has been analyzed by conducting integral tests at a scale-down test facility, named Purdue University Multi-Dimensional Integral Test Assembly (PUMA). The integral tests simulate the hypothetical Feed Water Line Break (FWLB) accident caused by a guillotine rupture of the Feed Water Line (FWL) attached to the reactor vessel. This research is motivated to evaluate the behaviors of passive safety systems and to provide experimental data for the Validation and Verification (V&V) of related system codes. The experiment simulates the LOCA transient after the reactor vessel depressurizes to 1 MPa (150 psi) from normal operating pressure. The early high-pressure blowdown process of LOCA transient is simulated by the system code. The plant system conditions at 1.0 MPa calculated by the system code are interpreted as the initial conditions for the integral test with appropriate scaling. The normal FWLB integral test is also repeated with the partial failure of the PCCS to evaluate the safety margin. Experimental data indicates that this LB LOCA is successfully mitigated by the functioning of Emergency Core Cooling System (ECCS) and Passive Containment Cooling system (PCCS), with no occurrence of core uncover or containment overpressure.

Evaluation of the asymmetry effect on core thermal limit for the inadvertent cold coolant injection transient in Chinshan BWR-4 Plant

The HPCI is designed to inject into one of the two feedwater lines, which eventually results in the uneven distribution of core inlet temperature. This paper develops a methodology to determine the penalty of HPCI flow to cover the defect of RETRAN computer code in the volume-averaged parameter and satisfy the requirement for a conservative analysis. The detailed inlet temperature calculated by CFD supports the determination of ΔCPR by using RETRAN hot channel model for the selected 56 potential channels.The Spearman rank correlation coefficient method is adopted to evaluate the dependence of the ΔCPR on the characteristics in the channel. Finally, the inlet subcooling in the channel is found to be in a medium to strong dependence on CPR change. The sensitivity studies verify that 20% penalty kept in HPCI flow covers 95% subcooling in all channels and the calculated ΔCPR has a rank of 54 (from low to high), which can cover 96.4% of CPR change in the 56 potential channels.

Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection.

Safety evaluation of the modified auxiliary feedwater system for the Chinese improved PWR

Chinese National Nuclear Safety Administration (NNSA) has issued the general technique requirements of the modification actions for the Nuclear Power Plants (NPPs). The enlargement of the tank volume of the auxiliary feedwater system (ASG) is one of the modification actions to enhance the safety margin for the Chinese improved PWR NPPs. Therefore, it is necessary to evaluate the tank volume of ASG can supply enough cooling water to the secondary side of the steam generators under the transient accidents, to maintain the reactor under safety situation. Typical accident sequences, such as the category II accidents of the loss of feedwater (LOFW) with no loss of offsite power (no-LOOP) and the category IV accidents of the feedwater line break (FWLB) with no-LOOP and the FWLB combined with LOOP, are selected to analyze the minimum flow rate and volume for ASG, based on the RCC-P requirements, based on the model of the typical improved generation II 1000MWe nuclear power plant, built by using the mechanical safety analytical code. The auxiliary feedwater flow rates of 5m3/h, 10m3/h, 15m3/h and 25m3/h to each steam generator (SG) for the LOFW with no-LOOP are investigated by analyzing the transient progress, and the results show that the minimum flow rate is 25m3/h. FWLB transients with different flow rates of 30m3/h, 35m3/h and 45m3/h are also analyzed to show that 45m3/h is the minimum flow rate considering a safety margin. The minimum volume of the auxiliary feedwater tank is 713m3 for the LOFW, 723m3 for the FWLB with no-LOOP and 799m3 for the FWLB combined with LOOP. Therefore, considering the safety margin, the minimum auxiliary feedwater flow rate necessary to prevent bulk boiling in the reactor coolant system is 45m3/h for each unbroken SG, and a 900-m3 volume of auxiliary feedwater tank can satisfy the RCC-P requirements for the improved PWR NPPs, to give more time to take mitigation measurements.

Accident analyses for china pressurizer reactor with an innovative conceptual design of passive residual heat removal system

An innovative passive safety system design concept for CPR 1000 nuclear power plant (NPP), proposed based on the original passive residual heat removal system (PRHRS) design, is presented and studied in this paper. The new type PRHRS avoids use of trigger valve inside the containment and adds an external water tank as the trip of PRHRS operation. The new design concept can realize functions of system remote control and trigger, while the external water tank can provide emergency water for steam generator (SG) in case of loss of feed water accident. Also, the advantages to use a valve and a tank located externally to containment are including the more reliability in implementing valve in pipes at low pressure, possibility operating manually the valve to start the system, refilling the tank in case of long time cooling and maintaining the valve easily. Best-estimate transient simulation code Relap5/MOD3.4 is applied to study the behavior of new design PRHRS and transient characteristics of primary loop system during normal condition and accident conditions. The transient processes of Station Black-Out (SBO) and Feed-water Line Break (FLB) accidents are studied to verify the function of new design PRHRS. Results show that the effect of valve position change from inside to outside containment can be ignored for the new design PRHRS in normal operation and the external valve can trip the operation of PRHRS in case of accidents. The new design PRHRS can remove core residual heat from primary loop effectively through establishing the stable natural circulations in the primary loop and PRHRS loop when the SBO and FLB accidents occur. It also can supply the emergency water to the SG in the event of FLB accident and realize safe reactor shutdown. Results indicate that the new design concept PRHRS in this work is potential and successful in future application.

Efficiency and power upgrade by an additional high pressure economizer installation at an aged 620 MWe lignite-fired power plant

An additional high pressure economizer was installed at Unit B1 of the 620 MWe lignite-fired Power Plant “Nikola Tesla B” after 30 years of its operation. An innovative connection of the new additional economizer was applied. It is in parallel connection to the first section of the originally built economizer and it is directly fed with the feedwater from the outlet of the feedwater pump. The analysis of Unit B1 operation with such an economizer arrangement is performed and it is supported by measured data. It is shown that more than 30 MWth of the flue gas waste heat is recovered. The Unit gross efficiency is increased by 0.53 percentage points, which leads to 9.4 MWe of the electric power production. The parallel connection of the additional economizer also leads to the partial feedwater bypass of the high pressure heaters, which enables an increase of the plant electric power by up to 24.5 MWe. The accompanying effects are the reduction of the pressure drop in the feedwater line and the economizers, which leads to the decrease of the energy consumption for the main feedwater pump operation. The applied solution is presented together with measured and calculated energy and economic benefits.

Radioactivity evaluation for Main Steam Line and Suppression Chamber of small type BWR

*We, The Ja pan Atomic Power Company, have already started preparatory tasks for decommissioning of Tsuruga Nuclear Power Station unit 1 (TS-1), a small type BWR. Radiological characterization of the plant is one of important task of these tasks. We have performed radioactivity distribution calculation as radiological characterization around the reactor in Primary Containment Vessel (PCV) and in Biological Shielding Wall (BSW) and have obtained reliable information. But we needed more information on radioactivity distribution in TS-1 Reactor Building to plan rational decommissioning planning. In this work, we extended calculation areas to around Main Steam Line (MS), around Feed Water Line (FW) and around Suppression Chamber (S/C), where we needed special treatments. Local neutron transport phenomena, such as neutron streaming through MS, FW and Vent Tube, is observed there. To simulate such neutron transport phenomena well, we performed the calculations using three-dimensional neutron transport calculation code named TORT, applying Coupling Calculation technique. Calculations showed that all materials in these areas could be treated as non-radioactive. In order to verify reliability of calculations, we also performed radioactivity measurements. And Comparison between calculations and measurements proved that calculations had quantitative reliability.

Assessment of passive residual heat removal system cooling capacity

Advanced passive PWR relies on passive safety systems to provide core cooling capacity and deal with design basis accidents and beyond design basis accidents. However, the passive safety system is lack of practical operating experience and their performance is heavily influenced by other systems. The cooling capacity of passive residual heat removal system (PRHR), which is designed to remove decay heat when normal heat removal approach is not available, requires specific assessment during different accidents. In this study, a detail model of advanced passive PWR, including Reactor Coolant System (RCS), simplified secondary side and Engineered Safety Features (ESF), has been built using mechanism accident analysis code. The plant transient has been simulated, and cooling capacity of PRHR been analyzed during loss of normal feedwater and main feedwater line rupture. Conservative assumptions were made specially based on different accident scenarios and one of the two fail-open valves arranged in parallel at the PRHR heat exchanger (HX) outlet line was assumed not open, as the worst single failure. The progress of the two accident sequence is calculated and the thermalhydraulic behavior of RCS is investigated and the main transient parameters are obtained, including primary side pressure, steam generator pressure, pressurizer water level. The cooling power and system response are calculated. The results show that PRHR, with CMT injection, can remove the decay heat from RCS to IRWST, keeping the pressures of RCS and steam generators remaining below 110 percent of the design values and the pressurizer overfilling is prevented. Sensitivity study has been performed to study the system resistance effects on the capacity of PRHR, which shows that increase in system resistance coefficient reduces the cooling capacity of PRHR.

Long-term pressure and temperature analysis and suppression pool mixing of Lungmen ABWR containment

Lungmen Nuclear Power Plant in Taiwan is a twin-unit Advanced Boiling Water Reactor (ABWR) plant. The design-basis accident for the ABWR containment is a double-ended feedwater line break accident, and the long-term containment responses for this accident are calculated using GOTHIC in this study. The wetwell space is modeled in one dimension either in multi dimensions to investigate the pool mixing effect. The suppression pool is cooled by the Residual Heat Removal (RHR) system in a long-term accident. The RHR pumps have suction flow from the pool bottom. The suction flow is cooled by the heat exchangers, and then returns to the pool via the RHR return line. The long-term containment response using GOTHIC agrees with the Lungmen FSAR results. Significant stratification in the wetwell is observed using the one-dimensional model. However, the multi-dimensional wetwell models predict that the suppression pool and the wetwell airspace can be well mixed by the internal convection. The multi-dimensional results also show that using the wetwell spray to return the cooled water can reduce long-term wetwell pressure.

The comparison of designed water-cooled and air-cooled passive residual heat removal system for 300 MW nuclear power plant during the feed-water line break scenario

The steam generator (SG) secondary water-cooled and air-cooled passive residual heat removal systems (PRHRSs) are proposed and designed for 300MW Pressurized Water Reactor (PWR). The RELAP5/MOD3.4 code is utilized to study the behavior of the PRHRSs and transient characteristics of primary loop system during Feed-water Line Break (FLB) accident. The characteristic comparison between water-cooled and air-cooled PRHRSs is also conducted in this study. The results show that both water-cooled and air-cooled PRHRSs can establish stable natural circulations in PRHRS loops, which realize the effective core decay heat removal from primary loop. Results illustrate that both water-cooled PRHRS and air-cooled PRHRS designed in the study have great significance for improving the inherent safety of 300MW nuclear power plant. However, the water-cooled PRHRS heat transfer ability is stronger in the initial time, while it becomes weaker as the system tank water evaporates in the later stage. In contrast, the air-cooled PRHRS, ensuring the main reactor operation parameters to decrease to a more safety value, is more effective than water-cooled PRHRS for long term cooling.

Study of Improvement in Boiler Efficiency through Incorporation of Additional Bank of Tubes in the Economiser

The major efficiency loss of a boiler is caused by the hot stack gases discharging to the atmosphere which is polluting the atmosphere and on other side Pollution Control Board is forcing the norms of Pollution levels in atmosphere. One of the most cost-effective ways of improving the efficiency of a high pressure steam boiler is to install an economizer on the boiler. An economizer is a heat exchanger, which transfers heat from the stack gases to the incoming feedwater. Typically, on a high pressure water tube boiler, the efficiency improvement with an economizer is 2 to 4%, depending on firing rate. On a high pressure fire tube boiler, the improvement is 2 to 3.5%, depending on boiler size and firing rate. The economizers are located in the boiler stack close to the stack gas outlet of the boiler. They may be supported from overhead or from the floor. A feedwater line, which serves the boiler, is piped to the unit. No additional feedwater control valves or stack gas dampers are required. Presently in NTPC stage II units there are banks of tubes in economizer. There was a proposal from management to add another bank of tubes in the economizer so that there will be control of pollutants coming out from boiler. We took that proposal as a task and made a detailed investigation of it. An investigation is conducted on the effect of performance of the boiler by incorporating the additional bank of tubes in the space below the lower bank of tubes. The main idea is to extract maximum amount of heat from the flue gases and increase the heat pick up rate of the feed water outlet into the boiler. By assuming the additional bank of tubes, various heat transfer calculations are done and the reduction in the flue gas out let temperature and increase in the feedwater outlet temperature are found out. Finally a comparison of efficiencies of the boiler is made between the existing unit and the unit when additional bank of tubes are assumed. At last, various pros and cons have been found out by conducting above investigation i.e. by incorporating additional bank of tubes in the economizer.

SEPARATE AND INTEGRAL EFFECT TESTS FOR VALIDATION OF COOLING AND OPERATIONAL PERFORMANCE OF THE APR+ PASSIVE AUXILIARY FEEDWATER SYSTEM

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS)

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for third-generation (GEN-III) nuclear power plants that are driven by passive systems, such as natural circulation, gravity, and resistance to high temperatures. Thus, South Korea has designed the Advanced Power Reactor Plus (APR+) with a two-loop PWR and 1500 MWe by adding passive safety features to the Advanced Power Reactor 1400 MWe (APR1400). The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. Therefore, in this paper, after introducing the characteristics of the PAFS and its design requirements, a performance analysis of the PAFS is performed for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB). For the analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used, and a MARS model is developed by adding the PAFS model to the existing APR1400 model. The analysis results show that the PAFS has enough capacity to remove decay heat under the postulated accident conditions. In addition, the adequacy of modified control logic for main steam isolation valve (MSIV) is validated by comparing the traditional control logic.

원전 격실에 대한 최적 침수분석 방법

본 논문은 원자력발전소의 대형탱크 또는 배관파단에 따른 격실의 침수분석을 수행함에 있어 최적평가방법을 개발하여 원전에 실제로 적용하는 방법에 관한 논문을 작성하는데 목적이 있다. 주급수관파단사고 분석을 위해 RETRAN 전산코드를 사용하였다. 유출수 질량유량을 계산하는데 있어서 주급수제어밸브가 계통설계에 의거 원자로정지 후 5.0초 만에 닫히는 것으로 모델링하여 분석하였다. 출력 70% 운전시 방출유량이 가장 높은 것으로 나타났다. 방출 질량유량을 가지고 침수위를 계산한 결과 주급수관 격실의 최대 침수위는 1.43m로서 이는 안전성기기가 설치된 위치보다 낮아 원전의 안전정지에 미치는 영향이 없는 것으로 나타났다. In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.

Simulation of Normal/Abnormal Operational Conditions of Integral Pressurized Water Reactor by Using RELAP5/MOD3.4

This paper is focused on the normal and abnormal operational states of small nuclear reactor (integral pressurized water reactor) by analysing with thermal hydraulic system code RELAP5/MOD3.4. For this purpose, the reactor considered is INSURE-100 with the power output of 100 MW. Concerning with the normal operational conditions, the steady-state results as well as load variation characteristics have been carried out and for the abnormal conditions, preliminary assessment of the reactor under two accident conditions i.e., loss of feed water and main steam line break have been researched and are well explained with the help of graphical analysis by using RELAP5 code. The research is also focused on the safety of the considered reactor, and for this purpose the reactor was design to be of pool-type reactor. The graphical results and the parameters that have been extracted from the RELAP5/MOD3.4 code show good agreement for the safe evaluation of the reactor.