PUMP: analog-hybrid reactor coolant hydraulic transient model

International Nuclear Information System (INIS)

Grandia, M.R.

1976-03-01

The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

International Nuclear Information System (INIS)

Sculthorpe, B.R.; Johnson, K.M.

1986-01-01

With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

Trends and experiences in reactor coolant pump motors

International Nuclear Information System (INIS)

Anon.

1980-01-01

A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

Two-phase coolant pump model of pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos; Freitas, R.L.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Verification Test of Hydraulic Performance for Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

2010-01-15

According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

International Nuclear Information System (INIS)

Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

2009-01-01

The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

Reactor coolant pump shaft seal behavior during blackout conditions

International Nuclear Information System (INIS)

Mings, W.J.

1985-01-01

The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

Operation diagnostics of the reactor coolant pumps in the Jaslovske Bohunice nuclear power plant, CSSR

International Nuclear Information System (INIS)

Bahna, J.; Jaros, I.; Oksa, G.

1990-01-01

The state of the art of the materials basis, the diagnostics methods used, organization of data collection and processing, and some results of routine and specific investigations concerned with diagnosis of the reactor coolant pump in the Jaslovske Bohunice NPP V-1 are presented. Some information is given about the reactor coolant pump monitor developed in the VUJE. (author)

Structural integrity analysis of reactor coolant pump flywheel(I)

International Nuclear Information System (INIS)

Kim, Young Jin

1986-01-01

A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1992-09-01

This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal

Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

International Nuclear Information System (INIS)

Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

2013-01-01

The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

Station blackout with reactor coolant pump seal leakage

International Nuclear Information System (INIS)

Evinay, A.

1993-01-01

The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

Condition monitoring of main coolant pumps, Dhruva

International Nuclear Information System (INIS)

Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

2002-01-01

Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

Reactor coolant pump monitoring and diagnostic system

International Nuclear Information System (INIS)

Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

1990-01-01

In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

Power supplyer for reactor coolant recycling pump

International Nuclear Information System (INIS)

Nara, Hiroshi; Okinaka, Yo.

1991-01-01

The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

International Nuclear Information System (INIS)

Meng Qinghu; Meng Qingfeng; Feng Wuwei

2012-01-01

Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

International Nuclear Information System (INIS)

Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

2015-01-01

At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

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

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

2011-01-01

SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

OpenAIRE

Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

2012-01-01

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

Feasibility study on the type of KALIMER coolant circulation pump

International Nuclear Information System (INIS)

Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

1997-07-01

The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

Automated surveillance of reactor coolant pump performance

International Nuclear Information System (INIS)

Gross, K.C.; Singer, R.M.; Humenik, K.E.

1992-01-01

An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

International Nuclear Information System (INIS)

Hayner, G.O.; Lubnow, T.; Clary, M.

1990-01-01

On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

Regulatory analysis for Generic Issue 23: Reactor coolant pump seal failure. Draft report for comment

Energy Technology Data Exchange (ETDEWEB)

Shaukat, S K; Jackson, J E; Thatcher, D F

1991-04-01

This report presents the regulatory/backfit analysis for Generic Issue 23 (GI-23), 'Reactor Coolant Pump Seal Failure'. A backfit analysis in accordance with 10 CFR 50.109 is presented in Appendix E. The proposed resolution includes quality assurance provisions for reactor coolant pump seals, instrumentation and procedures for monitoring seal performance, and provisions for seal cooling during off-normal plant conditions involving loss of all seal cooling such as station blackout. Research, technical data, and other analyses supporting the resolution of this issue are summarized in the technical findings report (NUREG/CR-4948) and cost/benefit report (NUREG/CR-5167). (author)

Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

International Nuclear Information System (INIS)

Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

1994-01-01

Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

International Nuclear Information System (INIS)

Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

2014-01-01

The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

Reactor coolant pump service life evaluation for current life cycle optimization and license renewal

International Nuclear Information System (INIS)

Doroshuk, B.W.; Berto, D.S.; Robles, M.

1990-01-01

This paper reports that as part of the plant life cycle management and license renewal program, Baltimore Gas and Electric Company (BG and E) has completed a service life evaluation of their reactor coolant pumps, funded jointly by EPRI and performed by ABB Combustion Engineering Nuclear Power. Two of the goals of the BG and E plant life cycle management and license renewal program, and of this current evaluation, are to identify actions which would optimize current plant operation, and ensure that license renewal remains a viable option. The reactor coolant pumps (RCPs) at BG and E's Calvert Cliffs Units 1 and 2 are Byron Jackson pumps with a diffuser and a single suction. This pump design is also used in many other nuclear plants. The RCP service life evaluation assessed the effect of all plausible age-related degradation mechanisms (ARDMs) on the RCP components. Cyclic fatigue and thermal embrittlement were two ARDMs identified as having a high potential to limit the service life of the pump case. The pump case is a primary pressure boundary component. Hence, ensuring its continued structural integrity is important

Coolant clean-up system in nuclear reactor

International Nuclear Information System (INIS)

Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

1987-01-01

Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

In-operation diagnostic system for reactor coolant pump

International Nuclear Information System (INIS)

Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

1996-01-01

A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

International Nuclear Information System (INIS)

Gu Yingbin

2012-01-01

Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Reactor coolant pump seal response to loss of cooling

International Nuclear Information System (INIS)

Graham, T.; Metcalfe, R.; Burchett, P.

2000-01-01

This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

Reactor coolant pump shaft seal stability during station blackout

International Nuclear Information System (INIS)

Rhodes, D.B.; Hill, R.C.; Wensel, R.G.

1987-05-01

Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries

Reactor coolant pump shaft seal stability during station blackout

Energy Technology Data Exchange (ETDEWEB)

Rhodes, D B; Hill, R C; Wensel, R G

1987-05-01

Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

International Nuclear Information System (INIS)

Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

2010-01-01

The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

International Nuclear Information System (INIS)

Vukovic, J.; Besirevic, A.; Boljat, Z.

2016-01-01

After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

Main coolant pump testing at Ontario Hydro

International Nuclear Information System (INIS)

Hartlen, R.

1991-01-01

This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

Secondary seal effects in hydrostatic non-contact seals for reactor coolant pump shaft

International Nuclear Information System (INIS)

Fujita, T.; Koga, T.; Tanoue, H.; Hirabayashi, H.

1987-01-01

The paper presents a seal flow analysis in a hydrostatic non-contact seal for a PWR coolant pump shaft. A description is given of the non-contact seal for the reactor coolant pump. Results are presented for a distortion analysis of the seal ring, along with the seal flow characteristics and the contact pressure profiles of the secondary seals. The results of the work confirm previously reported findings that the seal ring distortion is sensitive to the o-ring location (which was placed between the ceramic seal face and the seal ring retainer). The paper concludes that the seal flow characteristics and the tracking performance depend upon the dynamic properties of the secondary seal. (U.K.)

Coolant clean-up system in the primary coolant circuit for nuclear reactor

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

International Nuclear Information System (INIS)

Morris, D.J.; Gabler, H.C.

1982-10-01

Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

Reactor coolant pump shaft seal behavior during station blackout

International Nuclear Information System (INIS)

Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

1985-04-01

A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

Coast-down model based on rated parameters of reactor coolant pump

International Nuclear Information System (INIS)

Jiang Maohua; Zou Zhichao; Wang Pengfei; Ruan Xiaodong

2014-01-01

For a sudden loss of power in reactor coolant pump (RCP), a calculation model of rotor speed and flow characteristics based on rated parameters was studied. The derived model was verified by comparing with the power-off experimental data of 100D RCP. The results indicate that it can be used in preliminary design calculation and verification analysis. Then a design criterion of RCP was described based on the calculation model. The moment of inertia in AP1000 RCP was verified by this criterion. (authors)

Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

International Nuclear Information System (INIS)

Lee, K.B.; Im, I.Y.; Lee, S.K.

1992-01-01

An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

Triboengineering problems of lead coolant in innovative fast reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

2013-01-01

Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

Failures of the thermal barriers of 900 MWe reactor coolant pumps

International Nuclear Information System (INIS)

Peyrouty, P.

1997-01-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enables faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk represented by deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible. (author)

Failures of the thermal barriers of 900 MWe reactor coolant pumps

Energy Technology Data Exchange (ETDEWEB)

Peyrouty, P.

1996-12-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

Failures of the thermal barriers of 900 MWe reactor coolant pumps

International Nuclear Information System (INIS)

Peyrouty, P.

1996-01-01

This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

Energy Technology Data Exchange (ETDEWEB)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-04-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

International Nuclear Information System (INIS)

Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

1997-01-01

This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

1983-01-01

Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

Directory of Open Access Journals (Sweden)

Lefèvre Grégory

2012-01-01

Full Text Available In the primary circuit of pressurized water reactors (PWR, the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek theory and used as such to interpret this industrial phenomenon.

Qualification test of a main coolant pump for SMART pilot

International Nuclear Information System (INIS)

Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

2006-01-01

SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

Lubrication analysis of the journal bearing in the main coolant pump of SMART

International Nuclear Information System (INIS)

Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

2000-01-01

Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

Impedance calculations for power cables to primary coolant pump motors

International Nuclear Information System (INIS)

Hegerhorst, K.B.

1977-01-01

The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

Leak rate analysis of the Westinghouse Reactor Coolant Pump

International Nuclear Information System (INIS)

Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

1985-07-01

An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Speed control device for coolant recycling pump

International Nuclear Information System (INIS)

Kageyama, Takao.

1992-01-01

The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

Prediction of Hydraulic Performance of a Scaled-Down Model of SMART Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Kwon, Sun Guk; Park, Jin Seok; Yu, Je Yong; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2010-08-15

An analysis was conducted to predict the hydraulic performance of a reactor coolant pump (RCP) of SMART at the off-design as well as design points. In order to reduce the analysis time efficiently, a single passage containing an impeller and a diffuser was considered as the computational domain. A stage scheme was used to perform a circumferential averaging of the flux on the impeller-diffuser interface. The pressure difference between the inlet and outlet of the pump was determined and was used to compute the head, efficiency, and break horse power (BHP) of a scaled-down model under conditions of steady-state incompressible flow. The predicted curves of the hydraulic performance of an RCP were similar to the typical characteristic curves of a conventional mixed-flow pump. The complex internal fluid flow of a pump, including the internal recirculation loss due to reverse flow, was observed at a low flow rate.

On-line monitoring of main coolant pump seals

International Nuclear Information System (INIS)

Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

1984-06-01

The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

Diapo, applying advanced AI methods to diagnosis of PWR reactor coolant pump

International Nuclear Information System (INIS)

Porcheron, M.; Ricard, B.

1993-01-01

Electricite de France has decided to increase the capabilities of its monitoring and diagnostic architecture with the development of an AI system for reactor coolant pump diagnostic support. This development is carried out with the cooperation of the equipment constructor Jeumont Schneider Industries. This diagnostic system will eventually be included in an integrated surveillance architecture. We present the architecture of the system and the basics of the knowledge model used. Main data for diagnosis are provided by sensor data issued by the pump monitoring system. Diagnostic reasoning is based on the cooperation of two main activities : a heuristic search among typical symptomatic situations that leads to the formulation of hypotheses and a ''deep'' causal analysis that consists in backtracking from identified situations up to initial faults or causes. This approach is well fitted to field expert reasoning, and provides powerful diagnostic capabilities that help to overcome conventional limitations of expert systems entirely based on heuristic knowledge. (authors). 9 figs., 11 refs

Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

International Nuclear Information System (INIS)

Morris, D.J.; Sommerfield, G.A.

1983-12-01

The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

Inspection of the Sizewll 'B' reactor coolant pump flywheels

International Nuclear Information System (INIS)

McNulty, A.L.; Cheshire, A.

1992-01-01

The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

Lubrication analysis of the thrust bearing in the main coolant pump of SMART

International Nuclear Information System (INIS)

Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

2001-01-01

Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Brachet, Alain.

1981-01-01

This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

Statistical analysis of the Ft. Calhoun reactor coolant pump system

International Nuclear Information System (INIS)

Patel, Bimal; Heising, C.D.

1997-01-01

In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specification limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (Author)

Integrated main coolant pumps for pressurized-water reactors

International Nuclear Information System (INIS)

Wieser, R.

1975-01-01

The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

Coolant clean-up and recycle systems

International Nuclear Information System (INIS)

Ito, Takao.

1979-01-01

Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

Particle image velocimetry measurement of complex flow structures in the diffuser and spherical casing of a reactor coolant pump

Directory of Open Access Journals (Sweden)

Yongchao Zhang

2018-04-01

Full Text Available Understanding of turbulent flow in the reactor coolant pump (RCP is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms. Keywords: Diffuser, Flow Structures, Particle Image Velocimetry, Reactor Coolant Pump, Spherical Casing, Velocity Distribution

Statistical analysis of the Ft. Calhoun reactor coolant pump system

International Nuclear Information System (INIS)

Heising, Carolyn D.

1998-01-01

In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

International Nuclear Information System (INIS)

Azarm, M.A.; Boccio, J.L.; Mitra, S.

1985-12-01

This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs

Team training using full-scale reactor coolant pump seal mock-ups

International Nuclear Information System (INIS)

McDonald, T.J.; Hamill, R.W.

1987-01-01

The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

Application of damage function analysis to reactor coolant circuits

International Nuclear Information System (INIS)

MacDonald, D.D.

2002-01-01

The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

Application of damage function analysis to reactor coolant circuits

Energy Technology Data Exchange (ETDEWEB)

MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

2002-07-01

The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

Decontamination of main coolant pumps

International Nuclear Information System (INIS)

Roofthooft, R.

1988-01-01

Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

Analysis of Pressure Pulsation Induced by Rotor-Stator Interaction in Nuclear Reactor Coolant Pump

Directory of Open Access Journals (Sweden)

Xu Zhang

2017-01-01

Full Text Available The internal flow of reactor coolant pump (RCP is much more complex than the flow of a general mixed-flow pump due to high temperature, high pressure, and large flow rate. The pressure pulsation that is induced by rotor-stator interaction (RSI has significant effects on the performance of pump; therefore, it is necessary to figure out the distribution and propagation characteristics of pressure pulsation in the pump. The study uses CFD method to calculate the behavior of the flow. Results show that the amplitudes of pressure pulsation get the maximum between the rotor and stator, and the dissipation rate of pressure pulsation in impellers passage is larger than that in guide vanes passage. The behavior is associated with the frequency of pressure wave in different regions. The flow rate distribution is influenced by the operating conditions. The study finds that, at nominal flow, the flow rate distribution in guide vanes is relatively uniform and the pressure pulsation amplitude is the smallest. Besides, the vortex shedding or backflow from the impeller blade exit has the same frequency as pressure pulsation but there are phase differences, and it has been confirmed that the absolute value of phase differences reflects the vorticity intensity.

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

A dynamic model of the reactor coolant system flow for KMRR plant simulation

International Nuclear Information System (INIS)

Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

1990-01-01

To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

Study on the VFD (Variable Frequency Drive) for RCP (Reactor Coolant Pump) Motors of APR1400

Energy Technology Data Exchange (ETDEWEB)

Park, Jung Ha; Robert, M. Field; Kim, Tae Ryong [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

Most industrial facilities are continually searching for ways to reduce energy costs while increasing or maintaining current production. In terms of electric motors, Variable Frequency Drive (VFD) units represent a critical opportunity for energy savings. Currently, VFDs are used on about ten (10) percent of industrial process motors, and this percentage is increasing every year. Properly applied VFDs have been documented to save as much as fifty percent of the energy consumed by certain industrial processes. Nuclear Power - Power plants in general and Nuclear Power Plants (NPPs) in particular are slow to adopt new technology. The nuclear power industry requires a nearly absolute demonstration through operating experience in other industries in which the new approach will result in a net improvement in plant reliability without any surprises. Only recently has the nuclear industry begun to adapt VFD units for large motors. Specifically, there are several examples in the Boiling Water Reactor (BWR) fleet of replacing Motor-Generator (M-G) sets with VFD units for Reactor Recirculation (RR) pump motor service. At one station, VFD units were introduced upstream of the Circulating Water (CWP) pump motors to address environmental issues. They units are taking advantage of VFD technology whose benefits include increased reliability, reduction in electrical house load, improved flow control, and reduced maintenance. RCP Application - In the case of new generation, it has been reported that the Westinghouse AP1000 will make use of VFD units for the Reactor Coolant Pump (RCP) motors.

Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

International Nuclear Information System (INIS)

Hermansky, Jozef; Zavodsky, Martin

2014-01-01

A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

International Nuclear Information System (INIS)

Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

2009-01-01

In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

Coolant inlet device for nuclear reactors

International Nuclear Information System (INIS)

Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

1969-01-01

Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

Main-coolant-pump shaft-seal reliability investigation. Interim report

International Nuclear Information System (INIS)

Fair, C.E.; Marsi, J.A.; Greer, A.O.

1982-09-01

This report contains the results of a survey of reactor coolant pump shaft seal reliability. The survey sample is representatively large (approx. = 27% of total US commercial plant population) and includes the three industry seal suppliers (Bingham-Williamette, Byron Jackson, and Westinghouse). Operationally incurred/induced problems and seal redesign parameters are identified. Failure hypotheses in the form of fault trees have been developed to describe the failure mechanisms. Recommendations are made for seal reliability improvement

Development for LMR coolant technology - Development of a submersible-in-pool electromagnetic pump

Energy Technology Data Exchange (ETDEWEB)

Hong, Sang Hee; Kim, Hee Reyoung; Lee, Sang Don; Seo, Chun Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

1995-08-01

The conceptual and detailed designs of an annular linear induction electromagnetic pump of small scale submersible-in-pool type are performed for the purpose of domestic development of the pumps used for the high-temperature natrium coolant transportation in liquid metal reactors. The pump drawings for and input power of 1,100 VA, an input frequency of 17 Hz, a maximum flowrate of 60 l/min and a maximum operation temperature of 600 deg C are obtained from the optimum design analyses by solving MHD and equivalent circuit equations. The characteristics of pump materials in the high temperature and neutron irradiation environment are reflected in designing the pump, and theoretical analyses for improving the pump performance and efficiency are tried through calculations of magnetic flux and temperature distributions inside the pump. The present project contributes to the further design of engineering proto-type electromagnetic pump with higher capacity and the development of liquid metal reactor with innovative simplicity. 44 refs., 4 tabs., 33 figs. (author)

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Phenomena occuring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1990-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

Effect of the Shrink Fit and Mechanical Tolerance on Reactor Coolant Pump Flywheel Integrity Evaluation

Energy Technology Data Exchange (ETDEWEB)

Kim, Donghak [Korea KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Reactor coolant pump (RCP) flywheel should satisfy the RCP flywheel integrity criteria of the US NRC standard review plan (SRP) 5.4.1.1 and regulatory guide (RG) 1.14. Shrink-fit and rotational stresses should be calculated to evaluate the integrity. In this paper the effects of the shrink fit and mechanical tolerance on the RCP flywheel integrity evaluation are studied. The shrink fit should be determined by the joint release speed and the stresses in the flywheel will be increased by the shrink fit. The stress at the interface between the hub and the outer wheel shows the highest value. The effect of the mechanical tolerance should be considered for the stress evaluation. And the effect of the mechanical tolerance should be not considered to determine the joint release speed.

Effect of the Shrink Fit and Mechanical Tolerance on Reactor Coolant Pump Flywheel Integrity Evaluation

International Nuclear Information System (INIS)

Kim, Donghak

2015-01-01

Reactor coolant pump (RCP) flywheel should satisfy the RCP flywheel integrity criteria of the US NRC standard review plan (SRP) 5.4.1.1 and regulatory guide (RG) 1.14. Shrink-fit and rotational stresses should be calculated to evaluate the integrity. In this paper the effects of the shrink fit and mechanical tolerance on the RCP flywheel integrity evaluation are studied. The shrink fit should be determined by the joint release speed and the stresses in the flywheel will be increased by the shrink fit. The stress at the interface between the hub and the outer wheel shows the highest value. The effect of the mechanical tolerance should be considered for the stress evaluation. And the effect of the mechanical tolerance should be not considered to determine the joint release speed

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

Brenciaglia, G.; Gurevich, Y.; Liu, G.

1996-01-01

The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

Reactor coolant pump testing using motor current signatures analysis

Energy Technology Data Exchange (ETDEWEB)

Burstein, N.; Bellamy, J.

1996-12-01

This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

Reactor coolant cleanup device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

Design, construction and testing of replacement nuclear coolant pump stators to meet today's equipment reliability expectations

International Nuclear Information System (INIS)

Fostier, L.; Howell, D.

2005-01-01

The reliability expectations of equipment and components in today's nuclear power plant are much greater than three or more decades ago when nuclear plants were first constructed due to economic impact of a failure. Very few components in a pressurized water reactor plant can have as much impact of the plants capacity factor as a catastrophic failure of a reactor coolant pump winding. This paper describes the maintenance approach taken by one North American utility in attempt to preclude such failures. The paper will discuss the challenges of the reactor coolant pump application and the enhancements made in the winding design and construction by the supplier to address failure mechanisms so as to better meet present reliability expectations in accordance with dedicated specifications. The paper will also present the in-process and final testing requirements and limits imposed in an attempt to ensure quality of the machine windings, along with selected test results from the stators that have been designed and constructed to these specifications to date. (author)

The empirical intensity of PWR primary coolant pumps failure and repair

International Nuclear Information System (INIS)

Milivojevicj, S.; Riznicj, J.

1988-01-01

The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

Multi-state reliability for coolant pump based on dependent competitive failure model

International Nuclear Information System (INIS)

Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

2013-01-01

By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

Segmentation of turbo generator and reactor coolant pump vibratory patterns: a syntactic pattern recognition approach

International Nuclear Information System (INIS)

Tira, Z.

1993-02-01

This study was undertaken in the context of turbogenerator and reactor coolant pump vibration surveillance. Vibration meters are used to monitor equipment condition. An anomaly will modify the signal mean. At the present time, the expert system DIVA, developed to automate diagnosis, requests the operator to identify the nature of the pattern change thus indicated. In order to minimize operator intervention, we have to automate on the one hand classification and on the other hand, detection and segmentation of the patterns. The purpose of this study is to develop a new automatic system for the segmentation and classification of signals. The segmentation is based on syntactic pattern recognition. For the classification, a decision tree is used. The signals to process are the rms values of the vibrations measured on rotating machines. These signals are randomly sampled. All processing is automatic and no a priori statistical knowledge on the signals is required. The segmentation performances are assessed by tests on vibratory signals. (author). 31 figs

Cleaning device for recycling pump motor cooling system in nuclear reactor

International Nuclear Information System (INIS)

Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

1996-01-01

The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

International Nuclear Information System (INIS)

Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

2007-01-01

In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

Evaluation on Safety of Stainless Steels in Chemical Decontamination Process with Immersion Type of Reactor Coolant Pump for Nuclear Reactor

International Nuclear Information System (INIS)

Kim, Seong Jong; Han, Min Su; Jang, Seok Ki; Kim, Ki Joon

2011-01-01

Due to commercialization of nuclear power, most countries have taken interest in decontamination process of nuclear power plant and tried to develop a optimum process. Because open literature of the decontamination process are rare, it is hard to obtain skills on decontamination of foreign country and it is necessarily to develop proper chemical decontamination process system in Korea. In this study, applicable possibility in chemical decontamination for reactor coolant pump (RCP) was investigated for the various stainless steels. The stainless steel (STS) 304 showed the best electrochemical properties for corrosion resistance and the lowest weight loss ratio in chemical decontamination process with immersion type than other materials. However, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 431 was sporadically observed. The sizes of their pitting corrosion also increased with increasing cycle numbers

LMFBR with booster pump in pumping loop

International Nuclear Information System (INIS)

Rubinstein, H.J.

1975-01-01

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Integral forged pump casing for the primary coolant circuit of a nuclear reactor: Development in design, forging technology, and material

International Nuclear Information System (INIS)

Austel, W.; Korbe, H.

1986-01-01

Developments in the forging of large casings for primary circuit coolant pumps for light water reactors in Germany are demonstrated beginning with the multiple forging fabricated version and ending with the integral forged type. This version is the result of the joint efforts of the pump manufacturer and the forgemaster after a cost-gain evaluation and represents an optimum solution in view of its functional and economical performance and also considering the high requirements for mechanical-technological properties, including homogeneity of the material. The development from 22 NiMoCr 3 7/A 508 Class 2 to 20 MnMoNi 5 5/A 508 Class 3 and their optimization will be demonstrated. This development is based mainly on minimizing the sulfur content and on vacuum carbon deoxidation (VCD), which results in a reduction of the A-segregations, in improving fracture toughness and isotropy, and in the desired fine-grain structure

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

Energy Technology Data Exchange (ETDEWEB)

Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

2015-05-15

A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

Coolant monitoring systems for PWR reactors

International Nuclear Information System (INIS)

Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

1987-01-01

The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

Summary of failed reactor coolant pump rotating assembly experience at Crystal River Unit 3

International Nuclear Information System (INIS)

Hayner, G.O.; Clary, M.D.

1992-01-01

Four reactor coolant pump (RCP) rotating assemblies (shafts) have failed or have severely cracked during operation at the Crystal River Unit 3 (CR-3) Nuclear Power Plant. The two failed shafts removed from RCP-1A have been extensively examined. All of the RCP shafts (except the D shaft) were fabricated from UNS S66286 superalloy (Alloy A-286). The D shaft was fabricated from UNS S20910 (Alloy XM-19/Nitronic 50). Torsional strain gauge analysis was performed on the RCP-1A shaft during the 1990 refueling outage. This type of analysis has not been performed previously on an operating RCP. Several results were found including: (1) the primary components of alternating torsional stress during normal RCP operation are impeller vane pass and a sub-2X torsional resonance with maximum components of ∼±0.8 ksi; (2) a typical vane pass cycle is initiated by an abrupt unloading of the shaft followed by a reload past equilibrium and a damped return to equilibrium; (3) a higher (compared to normal four pump operation) alternating torsional stress range resulted from solo operation of RCP-1A at low temperature and pressure (normal startup conditions); (4) the 2/0 combination produced the highest mean torsional stresses and the lowest alternating stresses and (5) a startup of a secured RCP with three operating pumps results in significantly higher alternating stress than a cold startup. The root cause RCP failure mechanism appears to involve RCP startup sequence at CR-3, peculiarities that necessitate this sequence and complex shaft stresses just above or under the journal bearing. The 1986 impeller bolt failure is not considered to be a root cause effect. It was also determined that fatigue cracking has always been responsible for both shaft initiation and propagation mechanisms and cracking can occur independent of shaft material

AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

International Nuclear Information System (INIS)

SUBUDHI, M.; MORANTE, R.; LEE, A.D.

2002-01-01

The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

International Nuclear Information System (INIS)

Fletcher, C.D.; Callow, R.A.

1989-01-01

The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

International Nuclear Information System (INIS)

Duce, S.W.; Marley, A.W.; Freeman, A.L.

1989-12-01

Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

Liquid metal coolant disposal from UKAEA reactors at Dounreay

International Nuclear Information System (INIS)

Adam, E.R.

1997-01-01

As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

Emergency cooling system with hot-water jet pumps for nuclear reactors

International Nuclear Information System (INIS)

Reinsch, A.O.W.

1977-01-01

The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

Moment inertia pump analysis used in the Rsg-Gas primary coolant loop under lofa condition

International Nuclear Information System (INIS)

Sudarmono; Setiyanto; Dhandhang, P.; Dibyo, S.; Royadi

1998-01-01

The moment inertia of primary cooling system analysis under LOFA condition has been done. It is potentially one of limiting design constraints of the RSG-GAS safety because the coolant flow rate reduces very rapidly under LOFA condition due to the low inertia circulation pumps. If a loss of flow accident occurs, the mass flow will decrease rapidly and the heat transfer coefficient between cladding and coolant will also decreases. As a consequence the fuel and cladding temperature will increase. The whole core was represented by the 1/4 sector and divided into 19 subchannels and 40 axial nodes. In the present study, moment inertia of pump analysis for RSG-GAS reactor was performed with COBRA-IV-I subchannel code. As the DNB correlation, W-3 Correlation was selected for base case. The flow and power transients under pump trip accident were determined from experiments. The result above compared with the design data are 75 kg m 2 and 81 Kg m 2 respectively. The result shows that the RSG-GAS requires the inertia more than 75 kg m 2

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

BWR series pump recirculation system

International Nuclear Information System (INIS)

Dillmann, C.W.

1992-01-01

This patent describes a recirculation system for driving reactor coolant water contained in an annular downcomer defined between a boiling water reactor vessel and a reactor core spaced radially inwardly therefrom. It comprises a plurality of circumferentially spaced second pumps disposed in the downcomer, each including an inlet for receiving from the downcomer a portion of the coolant water as pump inlet flow, and an outlet for discharging the pump inlet flow pressurized in the second pump as pump outlet flow; and means for increasing pressure of the pump inlet flow at the pump inlet including a first pump disposed in series flow with the second pump for first receiving the pump inlet flow from the downcomer and discharging to the second pump inlet flow pressurized in the first pump

Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

International Nuclear Information System (INIS)

Lu, S.C.; Larder, R.A.; Ma, S.M.

1981-01-01

This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

Nuclear reactor apparatus

International Nuclear Information System (INIS)

Braun, H.E.; Bonnet, H.P.

1978-01-01

The reactor and its containment, instead of being supported on a solid concrete pad, are supported on a truss formed of upper and lower reinforced horizontal plates and vertical walls integrated into a rigid structure. The plates and walls from chambers within which the auxiliary components of the reactor, such as valves, pumping equipment and various tanks, are disposed. Certain of the chambers are also access passages for personnel, pipe chases, valve chambers and the like. In particular the truss includes an annular chamber. This chamber is lined and sealed by a corrosion-resistant liner and contains coolant and serves as a refueling cooling storage tank. This tank is directly below the primary-coolant conductor loops which extend from the reactor above the upper plate. The upper plate includes a sump connected to the tank through which coolant flows into the tank in the event of the occurrence of a loss-of-coolant accident. The truss extends beyond the containment and has chambers in the extending annulus. Pumps for circulating the coolant between the refueling coolant storage tank and the reactor are provided in certain of these chambers. The pumps are connected to the reactor by relatively short coolant conductors. Access to these pumps is readily afforded through hatches in the extending annulus

Method of suppressing the deposition of Co-60 to primary coolant pipeways in a nuclear reactor

International Nuclear Information System (INIS)

Hoshi, Michio; Tachikawa, Enzo; Goto, Satoshi; Sagawa, Chiaki; Yonezawa, Chushiro.

1987-01-01

Purpose: To suppress the deposition of Co-60 to primary coolant pipeways in a nuclear reactor. Method: To reduce the accumulation of Co-60 by causing chemical species of extremely similar chemical property with soluble Co-60 to be present together in coolants and replacing the deposition of Co-60 to the primary coolant pipeways in a nuclear reactor with that of the coexistent chemical spacies. Ni or Zn is used as the coexistet chemical spacies of similar chemical property with Co-60. The coexistent amount is from 5 to 10 times of the soluble Co-60 in the primary coolants. Ni or Zn solution adjusted with concentration is poured into and mixed with the coolants from a water feed source by using a high pressure constant volume pump. The amount of Co-60 taken into the pipeways caused by corrosion due to high temperature coolant is reduced to about 1/5 as compared with the case of Co-60 alone if 1 ppb of soluble Co-60 is present in water and 5 ppb of soluble Ni or Zn is added and, reduced to 1/12 if the amount of Ni or Zn is 10 ppb. (Kamimura, M.)

Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

2014-08-15

The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

International Nuclear Information System (INIS)

Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

1996-01-01

A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Leak detection device for reactor coolant

International Nuclear Information System (INIS)

Oshima, Koichiro.

1990-01-01

In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

International Nuclear Information System (INIS)

Hoehne, T.; Rhode, U.

2000-01-01

The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

International Nuclear Information System (INIS)

Singh, A.K.

1995-01-01

Reactor recirculation pumps at Tarapur Atomic Power Station (TAPS) are vertical, single stage centrifugal pumps having mechanical shaft seals and are driven by vertical mounted 3.3 kV, 3 phase, 1500 h.p. electric motors. During these years of operation TAPS has gained enough experience and expertise on the maintenance of reactor recirculation pumps which are dealt in this article. Failure of mechanical shaft seals, damage on pump carbon bearings, motor winding insulation failures and motor shaft damage have been the main areas of concern on recirculation pump. A detailed procedure step by step with component sketches has helped in eliminating errors during shaft seal assembly and installation. Pressure breakdown devices in seal assembly were rebuilt. Additional coolant water injection for shaft seal cooling was provided. These measures have helped in extending the reactor recirculation pump seal life. Pump bearing problems were mainly due to failure of anti-rotation pins and dowel pins of bearing assembly. These pins were redesigned and strengthened. Motor stator winding insulation failures were detected. Stator winding replacement program has been taken up on regular basis to avoid winding insulation failure due to aging. 3 refs., 2 tabs., 7 figs

Development of LMR Coolant Technology - Development of a submersible-in-pool electromagnetic pump

Energy Technology Data Exchange (ETDEWEB)

Hong, Sang Hi; Kim, Hee Reyoung; Lee, Sang Don; Seo, Joon Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyoungki University, Suwon (Korea, Republic of)

1997-07-15

A submersible-in-pool type annular linear induction pumps of 60 l/min and 200 l/min, and 600 deg C has been designed with optimum geometrical and operating values found from MHD and circuit analyses reflecting the high-temperature characteristics of pump materials. Through the characteristics analyses inside the narrow flow channel of electromagnetic pump, the distribution of the time-varying flow field is calculated, and magnetic flux and force density are evaluated by end effects of linear induction electromagnetic pump and the instability analyses are carried out introducing one-dimensional linear perturbation. Testing the pump with the flow rate of 60 l/min in the suitably manufactured loop system shows a flow rate of 58 l/min at an input power of 1,377 VA with 60Hz. The design of a scaled-up pump is further taken into account LMR coolant system requiring increased capacity, and a basic analysis is carried out on the pump of 40,000 l/min for KALIMER. The present project contributes to the further design of engineering prototype electromagnetic pumps with higher capacity and to the development of liquid metal reactor with innovative simplicity. 89 refs., 8 tabs., 45 figs. (author)

Coolant processing device for nuclear reactor

International Nuclear Information System (INIS)

Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

1981-01-01

Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

Nuclear reactor with coolant circulation pumps

International Nuclear Information System (INIS)

Peck, D.A.; Stolecki, W.E.

1975-01-01

Thermally induced movement of a pump or a heat exchanger in the primary circuit of a PWR is made possible by a suspension device. This device must however be, so rigid that it does not yield in cases of emergency. For this purpose, in the case of the pump a lower ring is provided carrying the pump by means of four columns. The columns are flexibly supported on the ring and a fixed constuction. Turned about 90% from these columns, two additional horizontal bars are flexibly mounted on the ring and on the motor housing of the pump as well as on the fixed construction. At the upper end of the motor housing, two shock absorbers are hinged in the same way. The joints are shaped as ball- and socket hinges. (DG) [de

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Iron crud supply device to reactor coolant

International Nuclear Information System (INIS)

Baba, Takao.

1993-01-01

In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

Simulations and field tests of a reactor coolant pump emergency start-up by means of remote gas units

International Nuclear Information System (INIS)

Omahen, P.; Gubina, F.

1992-01-01

The problem of the reactor coolant pump start-up in case of emergency by means of remote gas power plant units was analyzed. In this paper a simulation model is developed which enabled a detailed simulation of the transient process occurring at the start-up. The start-up of the RCP motor set was simulated in case of available one and two gas units. The field tests were performed and the measured variable values complied well with the simulation results. Two gas units have been determined as a safe start-up scheme of the RCP motor set considering for safety reasons accepted busbars and motor protection settings. A derived model for deep rotor bars was experimentally confirmed as effective means for the RCP motor set start-up transient simulation. Start-up procedures have been designed and adopted to the safety procedures of the Nuclear Power Plant Krsko

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Sloshing of coolant in a seismically isolated reactor

International Nuclear Information System (INIS)

Wu, T.S.; Guildys, J.; Seidensticker, R.W.

1988-01-01

During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

International Nuclear Information System (INIS)

Koenig, J.E.

1983-10-01

Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

International Nuclear Information System (INIS)

Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

1979-10-01

This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

International Nuclear Information System (INIS)

Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

2016-01-01

KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

International Nuclear Information System (INIS)

Andreychek, T.S.

1987-01-01

To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

Motor-pump unit provided with a lifting appliance of the motor

International Nuclear Information System (INIS)

Veronesi, Luciano; Francis, W.R.

1978-01-01

This invention relates to lifting appliances and particularly concerns a 'pump and motor set' or motor-pump unit fitted with a lifting appliance enabling the motor to be separated from the pump. In nuclear power stations the reactor discharges heat that is carried by the coolant to a distant point away from the reactor to generate steam and electricity conventionally. In order to cause the reactor coolant to flow through the system, coolant motor-pump units are provided in the cooling system. These units are generally of the vertical type with an electric motor fitted vertically on the pump by means of a cylindrical or conical structure called motor support [fr

Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-10-01

This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

Emergency recirculation pump driving mechanism

International Nuclear Information System (INIS)

Morooka, Shin-ichi.

1980-01-01

Purpose: To sufficiently secure the coolant flow rate in a reactor core and restrict the temperature on the surface of fuel elements to low degree when the coolant is lost in a BWR type reactor. Constitution: In order to secure sufficient coolant flow rate in a reactor core and to sufficiently cool the reactor core when the coolant is lost in a BWR type reactor, it is tripped upon loss of power supply simultaneously when an accident occurs, a recycling pump at the side of normal reactor where its rotating speed is decelerated in accordance with its inertia is restarted by the pressure water stored in a tank out of the reactor to increase the coolant flow rate in the reactor core so as to sufficiently cool the reactor core. (Aizawa, K.)

Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

Energy Technology Data Exchange (ETDEWEB)

Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

2016-11-15

The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

2015-10-15

In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

Consequences in the pumps operation during a large loss of coolant accident

International Nuclear Information System (INIS)

Santos, G.A. dos; Sabundjian, G.

1991-08-01

The event of living on or turning off the operation of the Reactor Cooling Pumps - RCPs, in the case of a Loss of Coolant Accident - LOCA, has been a reason of a lot of studies after the Three Mile Island 2 accident. Thus, it was investigated a large break LOCA in the cold leg of Angra 1, with the RELAP4/MOD5 Code during the blowdown. The attained results indicated that the best performance of the core was in the case where the RCPs had been turned off in the beginning of the transient, when compared with different operation conditions of the RCPs. (author)

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

Energy Technology Data Exchange (ETDEWEB)

Baek, Joo S.; Diamond, David

2016-12-06

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Residual heat removal pump and low pressure safety injection pump retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; McKenna, J.M.

1992-01-01

Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

RETRAN analysis of inter-system LOCA within the primary coolant pump

International Nuclear Information System (INIS)

Gangadharan, A.; Pratt, G.F.

1992-01-01

One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

Nuclear reactor

International Nuclear Information System (INIS)

Sakurai, Mikio; Yamauchi, Koki.

1983-01-01

Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

Continuous surveillance of reactor coolant circuit integrity

International Nuclear Information System (INIS)

1986-01-01

Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

Modeling the spatial distribution of the parameters of the coolant in the reactor volume

International Nuclear Information System (INIS)

Nikonov, S.P.

2011-01-01

In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts

International Nuclear Information System (INIS)

Bore, C.

1995-01-01

From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a 'viscosity pump' phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. In addition, final metallurgical

Magnetohydrodynamic generator and pump system

International Nuclear Information System (INIS)

Birzvalk, Yu.A.; Karasev, B.G.; Lavrentyev, I.V.; Semikov, G.T.

1983-01-01

The MHD generator-pump system, or MHD coupling, is designed to pump liquid-metal coolant in the primary circuit of a fast reactor. It contains a number of generator and pump channels placed one after another and forming a single electrical circuit, but hydraulically connected parallel to the second and first circuits of the reactor. All the generator and pump channels are located in a magnetic field created by the magnetic system with an excitation winding that is fed by a regulated direct current. In 500 to 2000 MW reactors, the flow rate of the coolant in each loop of the primary circuit is 3 to 6 m 3 /s and the hydraulic power is 2 to 4 MW. This paper examines the primary characteristics of an MHD generator-pump system with various dimensions and number of channels, wall thicknesses, coolant flow rates, and magnetic fields. It is shown that its efficiency may reach 60 to 70%. The operating principle of the MHD generator-pump system is explained in the referenced patent and involves the transfer of hydraulic power from generator channels to pump channels using a magnetic field and electrical circuit common to both channels. Variations of this system may be analyzed using an equivalent circuit. 7 refs., 5 figs

Method of injecting iron ion into reactor coolant

International Nuclear Information System (INIS)

Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

1988-01-01

Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

Radiolytic reactions in the coolant of helium cooled reactors

International Nuclear Information System (INIS)

Tingey, G.L.; Morgan, W.C.

1975-01-01

The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse

2012-01-01

the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

Reactor cooling apparatus

International Nuclear Information System (INIS)

Ogura, Kenji.

1983-01-01

Purpose: To increase natural convection flowrate in the reactor core upon interruption of a recycling pump by remarkably decreasing the flow resistance. Constitution: By-pass lines are disposed to a recycling pump in a primary coolant system and a second recycling pump in a secondary coolant system respectively, and a check valve and an isolation valve are attached to each of them. Each of the isolation valves is closed during normal operation and automatically opened when the number of rotation for each of the recycling pumps goes lower than a predetermined value. This can significantly decrease the flow resistance in the primary and secondary coolant systems upon interruption of the recycling pumps due to the entire loss of AC power source or the like to thereby increase the natural convection flowrate in the reactor core. (Sekiya, K.)

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

Condition monitoring of primary coolant pump-motor units of Indian PHWR

International Nuclear Information System (INIS)

Rshikesan, P.B.; Sharma, S.S.; Mhetre, S.G.

1994-01-01

As the primary coolant pump motor units are located in shut down accessible area, their start up, satisfactory operation and shut down are monitored from control room. As unavailability of one pump in standardised 220 MWe station reduces the station power to about 110 MWe, satisfactory operation of the pump is also important from economic considerations. All the critical parameters of pump shaft, mechanical seal, bearing system, motor winding and shaft displacement (vibrations) are monitored/recorded to ensure satisfactory operation of critical, capital intensive pump-motor units. (author). 2 tabs., 1 fig

Nuclear reactor recyclation device

International Nuclear Information System (INIS)

Takigawa, Yukio; Chuma, Kazuto

1987-01-01

Purpose: To prevent the unevenness for the coolant flow rate even when abnormality occurs to one of recycling pumps. Constitution: A plurality of jet pumps disposed at an interval around the reactor core are divided circumferentially into two sets, and a pipeway is disposed to the outside of each pair including recycling pumps corresponding to each of the sets. The pipeway is connected to the recycling inlet of the jet pump by way of a manifold. The discharge portion of the recycling pumps of the loop pipeway are connected with each other by way of communication pipes, and a normally closed valve is disposed to the communication pipe and the normally closed valve of the communication pipe is opened upon detecting abnormality for one of the recycling pumps. Thus, if either one of the pair of recycling pumps shows abnormal state, coolants flows from the other of pipeway to the outside of the loop pipeway and coolants are supplied from all the jet pumps to the reactor core portion and, accordingly, the not-uniform flow rate can be prevented to eliminate undesired effect on the reactor core. (Kamimura, M.)

Experimental and numerical investigation of the coolant mixing during fast deboration transients

International Nuclear Information System (INIS)

Hoehne, T.; Rohde, U.; Weiss, F.P.

1999-01-01

For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Probabilistic analyses of failure in reactor coolant piping

International Nuclear Information System (INIS)

Holman, G.S.

1984-01-01

LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

Full sized tests on a french coolant pump under two-phase flow

International Nuclear Information System (INIS)

Huchard, J.C.; Bore, C.; Dueymes, E.

1997-01-01

The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

Sodium as a reactor coolant

International Nuclear Information System (INIS)

Cesar, S.B.G.

1989-01-01

This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

Research on RCP400-TB50 type reactor coolant pump shaft seal failure analysis and monitoring method

International Nuclear Information System (INIS)

Yuan Chaolian; Shen Yuxian; Wang Chuan; Du Pengcheng

2014-01-01

Mechanical seal is widely applied in mechanical devices of nuclear power plant. 3-stages mechanical seal applied in reactor coolant pump (abbreviate to RCP) is a kind of product with top technology and manufacture difficulty. As the only running machine in primary loop of nuclear power plant, RCP is designed with high security, reliability and perform ability. So performance of its key component, 3-stages mechanical seal, could directly decide whether units can operate safely and reliably. In this paper mechanical seal used in RCP400-TB50 type RCP which in designed and manufactured by Andritz AG is selected as a typical example of dynamic pressure type mechanical seal applied in second generation NPP. Its structure and working principle is expounded. Engineering fluid mechanics theory is used to establish the mathematical model using for analyzing status of mechanical seal and deducing the theoretical formula. Its correctness is verified by compare with the test data. So that research result can be used as the theoretical basis for analysis of RCP400-TB50 RCP shaft seal's working condition. According to the shaft seal operation characteristic we can establish a suitable RCP shaft seal monitoring method and interlock protection setting for NPP operation. (authors)

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

1979-01-01

The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Conlon, W.M.; Lahey, R.T. Jr.

1982-12-01

This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

Recycling systems for BWR type reactors

International Nuclear Information System (INIS)

Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

1986-01-01

Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

National Research Council Canada - National Science Library

Can, Levent

2006-01-01

.... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

1998-03-01

The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

Nuclear reactor of pressurized liquid coolant type

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

Loss-of-coolant accident analysis of the Savannah River new production reactor design

International Nuclear Information System (INIS)

Maloney, K.J.; Pryor, R.J.

1990-11-01

This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

Pumps in nuclear power plants

International Nuclear Information System (INIS)

Kim, J.H.

1991-01-01

This paper reports that pumps play an important role in nuclear plant operation. For instance, reactor coolant pumps (RCPs) should provide adequate cooling for reactor core in both normal operation and transient or accident conditions. Pumps such as Low Pressure Safety Injection (LPSI) pump in the Emergency Core Cooling System (ECCS) play a crucial role during an accident, and their reliability is of paramount importance. Some key issues involved with pumps in nuclear plant system include the performance of RCP under two-phase flow conditions, piping vibration due to pump operating in two-phase flows, and reliability of LPSI pumps

Electromagnetic pump technology

International Nuclear Information System (INIS)

Prabhakar, R.

1994-01-01

Fast Breeder Reactors have an important role to play in our nuclear power programme. Liquid metal sodium is used as the coolant for removing fission heat generated in fast reactors and a distinctive physical property of sodium is its high electrical conductivity. This enables application of electromagnetic (EM) pumps, working on same principle as electric motors, for pumping liquid sodium. Due to its lower efficiency as compared to centrifugal pumps, use of EM pumps has been restricted to reactor auxiliary circuits and experimental facilities. As part of our efforts to manufacture fast reactor components indigenously, work on the development of two types of EM pumps namely flat linear induction pump (FLIP) and annular linear induction pump (ALIP) has been undertaken. Design procedures based on an equivalent circuit approach have been established and results from testing a 5.6 x 10E-3 Cum/s (20 Cum/h) FLIP in a sodium loop were used to validate the procedure. (author). 7 refs., 6 figs

Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

International Nuclear Information System (INIS)

Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

2014-01-01

The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

Structural evaluation of IEA-R1 primary system pump nozzles

Energy Technology Data Exchange (ETDEWEB)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel, E-mail: gfainer@ipen.br, E-mail: afaloppa@ipen.br, E-mail: calberto@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The IEA-R1 pumps of the primary coolant system may be required to withstand design and operational conditions. IEA-R1 nuclear research reactor is an open pool type reactor operated by IPEN since 1957. The reactor can be operated up to 5MW heating power since it was upgraded in a modernization program conducted by IPEN. The primary coolant system is composed by the piping system, decay tank, two heat pumps and two heat exchangers. In the latest arrangement upgrade of the primary system, conducted in 2014 as part of an aging management program, a partial replacement of the coolant piping and total replacement of piping and pump supports were done. As consequence, reviewed loads in the pump nozzles were obtained demanding a new evaluation of them. The aim of this report is to present the structural evaluation of the pump nozzles, considering the new loads coming from the new piping layout, according to: API 610 code verification, Supplier loads and structural analysis applying finite element method, by using the ANSYS computer program, regarding ASME VIII Div 1 & 2 recommendations. (author)

Structural evaluation of IEA-R1 primary system pump nozzles

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2017-01-01

The IEA-R1 pumps of the primary coolant system may be required to withstand design and operational conditions. IEA-R1 nuclear research reactor is an open pool type reactor operated by IPEN since 1957. The reactor can be operated up to 5MW heating power since it was upgraded in a modernization program conducted by IPEN. The primary coolant system is composed by the piping system, decay tank, two heat pumps and two heat exchangers. In the latest arrangement upgrade of the primary system, conducted in 2014 as part of an aging management program, a partial replacement of the coolant piping and total replacement of piping and pump supports were done. As consequence, reviewed loads in the pump nozzles were obtained demanding a new evaluation of them. The aim of this report is to present the structural evaluation of the pump nozzles, considering the new loads coming from the new piping layout, according to: API 610 code verification, Supplier loads and structural analysis applying finite element method, by using the ANSYS computer program, regarding ASME VIII Div 1 & 2 recommendations. (author)

Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report

International Nuclear Information System (INIS)

Fair, C.E.; Greer, A.O.

1983-03-01

This report presents a set of guidelines and criteria to aid in the generation of procurement specifications for Main Coolant Pump Shaft Seals. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies, a review of pump and shaft seal literature and discussions with pump and seal designers. This report is preliminary in nature and could be expanded and finalized subsequent to completion of further design, test and evaluation efforts

Internal pump monitoring device

International Nuclear Information System (INIS)

Kurosaki, Toshikazu.

1996-01-01

In the present invention, a thermometer is disposed at the upper end of an internal pump casing of a coolant recycling system in a BWR type reactor to detect leakage of reactor water thereby ensuring the improvement of reliability of the internal pump. Namely, a thermometer is disposed, which can detect temperature elevation occurred when water in the internal pump leaked from a reactor pressure vessel passes through the gap between a stretch tube and an upper end of the pump casing. Signals from the thermometer are transmitted to a signal processing device by an instrumentation cable. The signal processing device generates an alarm when the temperature signal exceeds a predetermined value and announces that leakage of reactor water occurs in the internal pump. Since the present invention can detect the leakage of the reactor water in the pump casing in an early stage, it can contribute to the improvement of the safety and reliability of the internal pump. (I.S.)

Residual heat removal pump retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; McKenna, J.M.

1990-01-01

Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor

International Nuclear Information System (INIS)

Nashine, B.K.; Rao, B.P.C.

2014-01-01

Highlights: • Derivation of applicable design equations. • Design of an annular induction pump based on these equations. • Testing of the designed pump in a sodium test facility. • Performance evaluation of the designed pump. - Abstract: Annular linear induction pumps (ALIPs) are used for pumping electrically conducting liquid metals. These pumps find wide application in fast reactors since the coolant in fast reactors is liquid sodium which a good conductor of electricity. The design of these pumps is usually done using equivalent circuit approach in combination with numerical simulation models. The equivalent circuit of ALIP is similar to that of an induction motor. This paper presents the derivation of equivalent circuit parameters using first principle approach. Sodium testing of designed ALIP using the equivalent circuit approach is also described and experimental results of the testing are presented. Comparison between experimental and analytical calculations has also been carried out. Some of the reasons for variation have also been listed in this paper

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

FBR type reactors

International Nuclear Information System (INIS)

Maemoto, Junko.

1985-01-01

Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

Analysis of loss-of-coolant accidents in pressurized water reactors

International Nuclear Information System (INIS)

Moldaschl, H.

1982-01-01

Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

Real-time analysis and display of reactor system mass inventory

International Nuclear Information System (INIS)

Dao, L.T.; Meachum, T.R.

1982-01-01

A mass inventory system (MIS) to evaluate, in real-time, the coolant distribution within the primary coolant system of the Loss-of-Fluid Test (LOFT) reactor has been developed. The computer-based system calculates and displays the coolant levels by two methods: using level measurements and performing a mass balance. The MIS is designed to provide up-to-date, intelligible information on the coolant distribution during any LOFT experiment. During LOFT experiments in which the primary coolant pumps are on, the method also provides void fraction information and the anticipated liquid level in the reactor vessel should the pumps be turned off

Fault diagnosis of main coolant pump in the nuclear power station based on the principal component analysis

International Nuclear Information System (INIS)

Feng Junting; Xu Mi; Wang Guizeng

2003-01-01

The fault diagnosis method based on principal component analysis is studied. The fault character direction storeroom of fifteen parameters abnormity is built in the simulation for the main coolant pump of nuclear power station. The measuring data are analyzed, and the results show that it is feasible for the fault diagnosis system of main coolant pump in the nuclear power station

Conceptual design of main coolant pump for integral reactor SMART

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Seok; Kim, Jong In; Kim, Min Hwan [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

The conceptual design for MCP to be installed in the integral reactor SMART was carried out. Canned motor pump was adopted in the conceptual design of MCP. Three-dimensional modeling was performed to visualize the conceptual design of the MCP and to check interferences between the parts. The theoretical design procedure for the impeller was developed. The procedures for the flow field and structural analysis of impeller was also developed to assess the design validity and to verify its structural integrity. A computer program to analyze the dynamic characteristics of the rotor shaft of MCP was developed. The rotational speed sensor was designed and its performance test was conducted to verify the possibility of operation. A prototypes of the canned motor was manufactured and tested to confirm the validity of the design concept. The MCP design concept was also investigated for fabricability by establishing the manufacturing procedures. 41 refs., 96 figs., 10 tabs. (Author)

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

Research on physical and chemical parameters of coolant in Light-Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

High-temperature process heat reactor with solid coolant and radiant heat exchange

International Nuclear Information System (INIS)

Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

1984-01-01

The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

Science.gov (United States)

Rezania, A.; Rosendahl, L. A.

2012-06-01

The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

Method and apparatus for removing radioactive gases from a nuclear reactor

International Nuclear Information System (INIS)

Frumerman, R.; Brown, W.W.

1975-01-01

A description is given of a method for removing radioactive gases from a nuclear reactor including the steps of draining coolant from a nuclear reactor to a level just below the coolant inlet and outlet nozzles to form a vapor space and then charging the space with an inert gas, circulating coolant through the reactor to assist the release of radioactive gases from the coolant into the vapor space, withdrawing the radioactive gases from the vapor space by a vacuum pump which then condenses and separates water from gases carried forward by the vacuum pump, discharging the water to a storage tank and supplying the separated gases to a gas compressor which pumps the gases to gas decay tanks. After the gases in the decay tanks lose their radioactive characteristics, the gases may be discharged to the atmosphere or returned to the reactor for further use

Study and application of boiling water reactor jet pump characteristic

International Nuclear Information System (INIS)

Liao Lihyih

1992-01-01

RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

Expanding the applicable duration for shrink fitting of the ultrathin-walled reactor coolant pump rotor-can

International Nuclear Information System (INIS)

Li, Ruiqin; Zhang, Chi; Zhang, Liwen; Cui, Yan; Shen, Wenfei

2017-01-01

Highlights: •A thermal-mechanical coupled finite element model was developed to simulate the whole process. •Heat capacity added layer was used to extend the limited time for the process. •Shrink-fitted experiments were performed to verify the simulation results. -- Abstract: The rotor-can of reactor coolant pump (RCP) is generally assembled on the rotor using shrink fitting technique. The rotor-can is characterized by large height and ultrathin-walled cylinder, thus, its rigidity is weak and heat capacity is quite limited. The shrink fitting process has to be completed within a short limited-time, which makes it difficult for rotor to insert in the rotor-can completely. In order to solve this problem, a new method was proposed to extend the limited time by using a heat capacity added layer (HCAL) during the shrink fitting process. A thermal-mechanical coupled finite element (FE) model was developed to simulate the whole process. The transient heat exchange with a narrow gap between rotor and rotor-can during the shrink fitting process was taken into consideration. The limited time was predicted by calculating and analyzing the evolutions of temperature field and radial displacement field of the rotor-can. The simulation results indicate that the limited time of the shrink fitting process can be significantly extended with the increase of HCAL in thickness. Then, shrink fitting experiments were performed to confirm the extending effect of the HCAL. The experimental results of limited time show good agreement with the predicted values. The current results will certainly help the designer to improve the shrink fitting technique.

Compatibility of structural materials with fusion reactor coolant and breeder fluids

International Nuclear Information System (INIS)

DeVan, J.H.

1979-01-01

Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

International Nuclear Information System (INIS)

Ware, A.G.

1994-07-01

Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

Fusion reactor pumped laser

International Nuclear Information System (INIS)

Jassby, D.L.

1988-01-01

A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

The solid coolant and prospects of its use in innovative reactors

International Nuclear Information System (INIS)

Dmitriev, A.M.; Deniskin, V.P.

2010-01-01

The progress of nuclear power demands consideration and development of innovative projects of the reactors having the increased level of safety due to their immanent properties allowing to provide high parameters. One of interesting and perspective offers is the use of a solid substance as a coolant. Use of the solid coolant of a nuclear reactor core has significant advantages among which an opportunity of movement of the coolant in the core under action of gravities and absence of necessity to have superfluous pressure in the jacket, that in turn means small metal consumption of construction, decrease in risk of emergency and its consequences. Cooling of the core with the help of solid substance is possible at performance of the certain conditions connected to features of the solid coolant. The major requirements are: the uniform continuous movement and minimal fluctuation of its density on every site of the core; high mechanical durability and wear resistance of particles; as well as good parameters of heat exchange, i.e. high heat conductivity and thermal capacity of the coolant material at the core operating conditions

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

International Nuclear Information System (INIS)

Werry, E.V.; Somasundaram, S.

1995-09-01

The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process

Measuring device for the coolant flowrate in a reactor core

International Nuclear Information System (INIS)

Sawa, Toshihiko.

1983-01-01

Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

N13 - based reactor coolant pressure boundary leakage system

International Nuclear Information System (INIS)

Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

1980-05-01

A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

Automatic coolant flow control device for a nuclear reactor assembly

Science.gov (United States)

Hutter, Ernest

1986-01-01

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Device for preventing coolant in a reactor from being lost

International Nuclear Information System (INIS)

Maruyama, Hiromi; Matsumoto, Tomoyuki.

1975-01-01

Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

International Nuclear Information System (INIS)

Bredikhin, V.Ya.; Moskvin, L.N.

1982-01-01

The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

International Nuclear Information System (INIS)

Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

1995-01-01

Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

Expert system for online surveillance of nuclear reactor coolant pumps

International Nuclear Information System (INIS)

Gross, K.C.; Singer, R.M.; Humenik, K.E.

1993-01-01

An expert system for determining the operability of a specified pump is described comprising: a set of pumps of which the specified pump is a member; means for measuring physical parameters representative to the operations condition each pump of said set of pumps; means for acquiring data generated by said measuring means; an artificial-intelligence based inference engine coupled to said data acquiring means where said inference engine applies a sequential probability ratio test to statistically evaluate said acquired data to determine a status for the specified pump and its respective measuring means by continually monitoring and comparing changes in a specific operational parameter signal acquired from a plurality of measurement means; means for transferring said status generated by said interference engine to an output system

Reactor feedwater pump control device

International Nuclear Information System (INIS)

Nishiyama, Hiroyuki.

1990-01-01

An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

International Nuclear Information System (INIS)

Piping Supriatna

2009-01-01

The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

LWR type reactor

International Nuclear Information System (INIS)

Kato, Kiyoshi.

1993-01-01

A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

International Nuclear Information System (INIS)

Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

1981-08-01

This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

Fracture assessment of a main reactor coolant pump in a BWR with encountered defects

Energy Technology Data Exchange (ETDEWEB)

Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

1988-12-31

This document presents a case-study fracture assessment in BWR type reactor components. A cast stainless steel presenting defects due to thermal is studied. The stress analysis performed by aid of a finite element technique shows that a Leak Before Break situation could be expected. Nevertheless, it may be concluded that the cross section of the pump where the defect area was located can withstand very deep cracks before the risk of failure becomes unacceptable. (TEC).

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

International Nuclear Information System (INIS)

Elliot, A.J.; Stuart, C.R.

2008-06-01

An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

The failure diagnoses of nuclear reactor systems

International Nuclear Information System (INIS)

Sheng Huanxing.

1986-01-01

The earlier period failure diagnoses can raise the safety and efficiency of nuclear reactors. This paper first describes the process abnormality monitoring of core barrel vibration in PWR, inherent noise sources in BWR, sodium boiling in LMFBR and nuclear reactor stability. And then, describes the plant failure diagnoses of primary coolant pumps, loose parts in nuclear reactors, coolant leakage and relief valve location

Coolant leakage detection device

International Nuclear Information System (INIS)

Ito, Takao.

1983-01-01

Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

Reactor coolant and associated systems in nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

Study of an electromagnetic pump in a sodium cooled reactor. Design study of secondary sodium main pumps (Joint research)

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru; Uchida, Akihito; Nishiguchi, Youhei; Nibe, Nobuaki

2006-07-01

In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750 MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160 m 3 /min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160 m 3 /min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160 m 3 /min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R and D plan of EMP are reported. (author)

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

2000-01-01

The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

Method of eliminating cruds in the primary coolants of reactors

International Nuclear Information System (INIS)

Tamura, Takaaki.

1984-01-01

Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

Fuel cladding interaction with water coolant in power reactors

International Nuclear Information System (INIS)

1985-11-01

Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

Thermodynamic analysis of the use a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water-splitting cycle for hydrogen production

International Nuclear Information System (INIS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

2008-01-01

Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved 'steam' parameters (outlet temperatures up to 625degC and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600degC. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the 'nuclear' heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of

Always at the correct temperature. Thermal management with electric coolant pump; Immer richtig temperiert. Thermomanagement mit elektrischer Kuehlmittelpumpe

Energy Technology Data Exchange (ETDEWEB)

Genster, A.; Stephan, W. [Pierburg GmbH, Neuss (Germany)

2004-11-01

Through the use of the electric coolant pump it has become possible for the first time to attain a cooling performance which is adapted precisely to the engine load and which is independent of engine speed. For cooling the new BMW six cylinder in-line Otto engine with an engine power rating of 190 kW, the electric coolant pump by Pierburg requires only 200 W of electrical power from the onboard electrical system. (orig.)

Sound velocity in the coolant of boiling nuclear reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

Modal method for crack identification applied to reactor recirculation pump

International Nuclear Information System (INIS)

Miller, W.H.; Brook, R.

1991-01-01

Nuclear reactors have been operating and producing useful electricity for many years. Within the last few years, several plants have found cracks in the reactor coolant pump shaft near the thermal barrier. The modal method and results described herein show the analytical results of using a Modal Analysis test method to determine the presence, size, and location of a shaft crack. The authors have previously demonstrated that the test method can analytically and experimentally identify shaft cracks as small as five percent (5%) of the shaft diameter. Due to small differences in material property distribution, the attempt to identify cracks smaller than 3% of the shaft diameter has been shown to be impractical. The rotor dynamics model includes a detailed motor rotor, external weights and inertias, and realistic total support stiffness. Results of the rotor dynamics model have been verified through a comparison with on-site vibration test data

Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

International Nuclear Information System (INIS)

1986-11-01

This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

International Nuclear Information System (INIS)

Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

2014-01-01

The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

Experiences in design up-gradation of mechanical seal cooling scheme of Dhruva PHT pumps

International Nuclear Information System (INIS)

Balakrishnan, K.T.P.

2002-01-01

Full text: Dhruva is a natural uranium fuelled high flux research reactor. Heavy water is used as coolant, moderator and reflector. Heat from the heavy water coolant is removed in heat exchangers by demineralised water. The heavy water coolant is re-circulated between the reactor core and the heat exchangers in three separate loops by three main coolant pumps (MCPs). The MCPs are high capacity centrifugal pumps and are rated for continuous service. The mechanical seal of the pump prevents leakage of the process fluid, which is heavy water, through the pump shaft. Continuous operation of the pump results in the heating up of the seal and necessitates sustained cooling. An integral cooling provision is made by tapping a 15 NB line from the discharge volute of the pump and feeding the process fluid itself as coolant to the seal. A non-indicating type flow-sensing device monitors flow through this line. Limiting values of flow are set and annunciated by a pair of magnetic reed type relays. This cooling line was a built in feature of the pumps as supplied by the manufacturer. This arrangement had the following inherent limitations: 1. There was no on line indication of the coolant flow. 2. The reed type magnetic relays initiated pump trips by spurious actuation, resulting in the interruption of reactor operation. Servicing a faulty flow switch involved lengthy procedures and necessitated draining, filling and venting of the pump. This entailed extended plant outages. Close proximity of these flow switches to a highly radioactive piping element imposed severe restrictions on the planned maintenance activity on them. Efforts were made to provide a suitable alternate cooling and flow measurement scheme to overcome the above-mentioned limitations. After evaluating the relative merits and demerits of several schemes, a turbine type flow sensor, on a modified cooling line was selected as the most suitable alternative. The alternate seal-cooling scheme was implemented for all

Experiences in design up-gradation of mechanical seal cooling scheme of Dhruva PHT pumps

International Nuclear Information System (INIS)

Balakrishnan, K.T.P.; Bharathan, R.

2002-01-01

Full text: Dhruva is a natural uranium fuelled high flux research reactor. Heavy water is used as coolant, moderator and reflector. Heat from the heavy water coolant is removed in heat exchangers by demineralised water. The heavy water coolant is re-circulated between the reactor core and the heat exchangers in three separate loops by three main coolant pumps (MCPs). The MCPs are high capacity centrifugal pumps and are rated for continuous service. The mechanical seal of the pump prevents leakage of the process fluid, which is heavy water, through the pump shaft. Continuous operation of the pump results in the heating up of the seal and necessitates sustained cooling. An integral cooling provision is made by tapping a 15 NB line from the discharge volute of the pump and feeding the process fluid itself as coolant to the seal. A non-indicating type flow-sensing device monitors flow through this line. Limiting values of flow are set and annunciated by a pair of magnetic reed type relays. This cooling line was a built in feature of the pumps as supplied by the manufacturer. This arrangement had the following inherent limitations : 1. There was no on line indication of the coolant flow. 2. The reed type magnetic relays initiated pump trips by spurious actuation, resulting in the interruption of reactor operation. Servicing a faulty flow switch involved lengthy procedures and necessitated draining, filling and venting of the pump. This entailed extended plant outages. Close proximity of these flow switches to a highly radioactive piping element imposed severe restrictions on the planned maintenance activity on them. Efforts were made to provide a suitable alternate cooling and flow measurement scheme to overcome the above-mentioned limitations. After evaluating the relative merits and demerits of several schemes, a turbine type flow sensor, on a modified cooling line was selected as the most suitable alternative. The alternate seal-cooling scheme was implemented for all

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Simulation of coolant mixing in pressure vessel reactors

International Nuclear Information System (INIS)

Hoehne, T.

2003-06-01

The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

Coolant mixing in pressurized water reactors. Proceedings

International Nuclear Information System (INIS)

Hoehne, T.; Grunwald, G.; Rohde, U.

1998-10-01

For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2017-08-15

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

International Nuclear Information System (INIS)

Boardman, J.R.

1994-10-01

This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

International Nuclear Information System (INIS)

Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

2003-01-01

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

Early detection of coolant boiling in research reactors with MTR-type fuel

International Nuclear Information System (INIS)

Kozma, R.; Turkcan, E.; Verhoef, J.P.

1992-10-01

A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

Boiling water reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

1975-01-01

Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

Water quality control device and water quality control method for reactor primary coolant system

International Nuclear Information System (INIS)

Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

1995-01-01

The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

Steam generator for a pressurized-water coolant nuclear reactor

International Nuclear Information System (INIS)

Schroeder, H.J.; Berger, W.

1975-01-01

A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

Transient simulation of coolant peak temperature due to prolonged fan and/or water pump operation after the vehicle is keyed-off

Science.gov (United States)

Pang, Suh Chyn; Masjuki, Haji Hassan; Kalam, Md. Abul; Hazrat, Md. Ali

2014-01-01

Automotive designers should design a robust engine cooling system which works well in both normal and severe driving conditions. When vehicles are keyed-off suddenly after some distance of hill-climbing driving, the coolant temperature tends to increase drastically. This is because heat soak in the engine could not be transferred away in a timely manner, as both the water pump and cooling fan stop working after the vehicle is keyed-off. In this research, we aimed to visualize the coolant temperature trend over time before and after the vehicles were keyed-off. In order to prevent coolant temperature from exceeding its boiling point and jeopardizing engine life, a numerical model was further tested with prolonged fan and/or water pump operation after keying-off. One dimensional thermal-fluid simulation was exploited to model the vehicle's cooling system. The behaviour of engine heat, air flow, and coolant flow over time were varied to observe the corresponding transient coolant temperatures. The robustness of this model was proven by validation with industry field test data. The numerical results provided sensible insights into the proposed solution. In short, prolonging fan operation for 500 s and prolonging both fan and water pump operation for 300 s could reduce coolant peak temperature efficiently. The physical implementation plan and benefits yielded from implementation of the electrical fan and electrical water pump are discussed.

Gear-shaft linkage, especially for nuclear reactor coolant pumps

International Nuclear Information System (INIS)

Delaunois, T.; Lefevre, R.

1990-01-01

The pump comprises: - inlet and outlet channels for the pumped fluid - a rotating shaft - a gear wheel mounted on the shaft by an axial locking nut which can support the axial hydraulic force - a thermal barrier above the gear wheel. A hydrostatic bearing fitted to the exterior surround of the gear wheel, the gear shaft linkage is made by at least a centering and locating device having a cylindrical span and an axial stop and another independent device which can take up the torque [fr

Coolant void effect investigation - case of a na-cooled fast reactor

International Nuclear Information System (INIS)

Glinatsis, G.; Gugiu, D.

2013-01-01

In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

Heat Radiators for Electromagnetic Pumps

Science.gov (United States)

Campana, R. J.

1986-01-01

Report proposes use of carbon/carbon composite radiators in electromagnetic coolant pumps of nuclear reactors on spacecraft. Carbon/carbon composite materials function well at temperatures in excess of 2,200 K. Aluminum has melting temperature of only 880 K.

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Detection of coolant void in lead-cooled fast reactors

International Nuclear Information System (INIS)

Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

2015-01-01

Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

Method of detecting coolant leakages from the pipeways in FBR type reactors

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1986-01-01

Purpose: To detect coolant leakage in the incore pipeways of loop type FBR type reactors in the initial stage at high sensitivity. Constitution: Temperature of the coolants sealed between incore pipeways and the buffle surrounding them is measured by thermocouples and coolant leakage is detected due to fluctuating components. A well-insertion type in which electrode is sealed with argon is used as the thermo-couples. Signals from the thermocouples are once amplified, removed with DC components and then only the fluctuating components are outputted. The fluctuating components are digitalized, passed through an adaptive digital filter and the RMS value as the difference between the output signal and the thermocouple signal is calculated. The calculated value is compared with a threshold value in a comparative calculator. If it exceeds the threshold value, it is judged as abnormal to display an alarm on an alarm display. In this way, the coolant leakage for the pipeways in the FBR type reactor can be detected on real time and at high sensitivity. (Kamimura, M.)

International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

Science.gov (United States)

Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

2011-01-01

The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

International Nuclear Information System (INIS)

Sun Yanfei; Zhong Shuoping; Huang Xiaojin

2014-01-01

Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

Radioactive corrosion products in circuit of fast reactor loop with dissociating coolant

International Nuclear Information System (INIS)

Dolgov, V.M.; Katanaev, A.O.

1982-01-01

The results of experimental investigation into depositions of radionuclides of corrosion origin on the surfaces of a reactor-in-pile loop facility with a dissociating coolant are presented. It is stated that the ratio of radionuclides in fixed depositions linearly decreases with decrease of the coolant temperature at the core-condenser section. The element composition of non-fixed compositions quantitatively and qualitatively differs from the composition of structural material, and it is more vividly displayed for the core-condenser section. The main mechanism of circuit contamination with radioactive corrosion products is substantiated: material corrosion in the zones of coolant phase transfer, their remove by the coolant in the core, deposition, activation and wash-out by the coolant from the core surfaces

Pump Coastdown with the Submerged Flywheel

Energy Technology Data Exchange (ETDEWEB)

Yoon, Hyun-Gi; Seo, KyoungWoo; Kim, Seong Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Many research reactors are generally designed as open pool types in consideration of the heat removal of the nuclear fuels, reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom as shown in Fig. 1. Primary cooling system pump circulates the coolant from the reactor structure to the heat exchanger in order to continuously remove the heat generated from the reactor core in the research reactor as shown in Fig. 1. The secondary cooling system releases the transferred heat to the atmosphere by the cooling tower. Coastdown flow rate of the primary cooling system pump with the submerged flywheel are calculated analytically in case of the accident situation. Coastdown flow rate is maintained until almost 80 sec when the pump stops normally. But, coastdown flow rate is rapidly decreased when the flywheel is submerged because of the friction load on the flywheel surface.

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

International Nuclear Information System (INIS)

Gonzalez, Jose; Gimenez, Marcelo

2000-01-01

We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

Rapid thermal transient in a reactor coolant channel

International Nuclear Information System (INIS)

Cherubini, A.

1986-01-01

This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

Energy Technology Data Exchange (ETDEWEB)

Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

2014-12-15

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

International Nuclear Information System (INIS)

Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

2014-01-01

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

Natural circulation in reactor coolant system

International Nuclear Information System (INIS)

Han, J.T.

1987-01-01

Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

Nuclear reactors

International Nuclear Information System (INIS)

Humphreys, P.; Davidson, D.F.; Thatcher, G.

1980-01-01

The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

Thermosyphoning in the CANDU reactor

International Nuclear Information System (INIS)

Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

1984-01-01

Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

2010-07-01

The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

Vacuum pumping for controlled thermonuclear reactors

International Nuclear Information System (INIS)

Watson, J.S.; Fisher, P.W.

1976-01-01

Thermonuclear reactors impose unique vacuum pumping problems involving very high pumping speeds, handling of hazardous materials (tritium), extreme cleanliness requirements, and quantitative recovery of pumped materials. Two principal pumping systems are required for a fusion reactor, a main vacuum system for evacuating the torus and a vacuum system for removing unaccelerated deuterium from neutral beam injectors. The first system must pump hydrogen isotopes and helium while the neutral beam system can operate by pumping only hydrogen isotopes (perhaps only deuterium). The most promising pumping techniques for both systems appear to be cryopumps, but different cryopumping techniques can be considered for each system. The main vacuum system will have to include cryosorption pumps cooled to 4.2 0 K to pump helium, but the unburned deuterium-tritium and other impurities could be pumped with cryocondensation panels (4.2 0 K) or cryosorption panels at higher temperatures. Since pumping speeds will be limited by conductance through the ducts and thermal shields, the pumping performance for both systems will be similar, and other factors such as refrigeration costs are likely to determine the choice. The vacuum pumping system for neutral beam injectors probably will not need to pump helium, and either condensation or higher temperature sorption pumps can be used

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

Science.gov (United States)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

New cooling system of the FRG-1 two barrier system of the primary coolant cycle

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2003-01-01

The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

Method for pre-heating lmfbr type reactors

International Nuclear Information System (INIS)

Yokozawa, Atsushi; Kataoka, Hajime.

1978-01-01

Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

International Nuclear Information System (INIS)

Aanstad, O.J.; Sklencar, A.M.

1983-01-01

A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

Core performance of equilibrium fast reactors for different coolant materials and fuel types

International Nuclear Information System (INIS)

Mizutani, Akihiko; Sekimoto, Hiroshi

1998-01-01

Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

Organic coolants and their applications to fusion reactors

International Nuclear Information System (INIS)

Gierszewski, P.; Hollies, B.

1986-08-01

Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

The flow measurement methods for the primary system of integral reactors

International Nuclear Information System (INIS)

Lee, J.; Seo, J. K.; Lee, D. J.

2001-01-01

It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

International Nuclear Information System (INIS)

Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

2000-01-01

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

A device for monitoring the coolant in a nuclear reactor tank

International Nuclear Information System (INIS)

Smith, R.D.

1984-01-01

The invention deals with a gamma thermometer where the gamma absorber (stainless steel) is in heat conducting connection with an external casing located in the coolant in a reactor tank. A heat sink for the gamma absorber heated by gamma irradiation from reactor fuel is thereby established. The most sensitive joint in the thermocouple of the gamma thermometer is mounted vertically above the other joint. A differential voltage with a certain polarity will be generated between the two joints during uniform cooling of the external casing. If the coolant falls to a level under the most sensitive joint, the resulting polarity change can be utilized for the activation of alarm systems. The same gamma thermometer may simultaneously be used as a sensor for measurement of local power distribution

Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

An optimized design of rectangle pumping cell for nuclear reactor pumped laser

International Nuclear Information System (INIS)

Wan, J.-S.; Chen, L.-X.; Zhao, Z.-M.; Pan, X.-B.; Jing, C.-Y.; Zhao, X.-Q.; Liu, F.-H.

2003-01-01

Basing on our research of energy deposition in RPL (Reactor Pumped Laser) pumping cell and the laser power efficiency, a RPL test device on Pulsed Reactor has been designed. In addition, the laser beam power of the RPL test device is estimated in the paper. (author)

The sodium coolant

International Nuclear Information System (INIS)

Rodriguez, G.

2004-01-01

The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

1991-01-01

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

Literature survey, numerical examples, and recommended design studies for main-coolant pumps. Final report

International Nuclear Information System (INIS)

Allaire, P.E.; Barrett, L.E.

1982-06-01

This report presents an up-to-date literature survey, examples of calculations of seal forces or other pump properties, and recommendations for future work pertaining to primary coolant pumps and primary recirculating pumps in the nuclear power industry. Five main areas are covered: pump impeller forces, fluid annuli, bearings, seals, and rotor calculations. The main conclusion is that forces in pump impellers is perhaps the least well understood area, seals have had some good design work done on them recently, fluid annuli effects are being discussed in the literature, bearing designs are fairly well known, and rotor calculations have been discussed widely in the literature. It should be noted, however, that usually the literature in a given area is not applied to pumps in nuclear power stations. The most immediate need for a combined theoretical and experimental design capability exists in mechanical face seals

Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

International Nuclear Information System (INIS)

Garrido, Oriol Costa; Cizelj, Leon; Simonovski, Igor

2012-01-01

Highlights: ► A 3-D model of a reactor coolant system for fatigue usage assessment. ► The performed simulations are a heat transfer and stress analyses. ► The main results are the expected ranges of fatigue loadings. - Abstract: The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

2012-09-01

To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

An introduction to the engineering of fast nuclear reactors

CERN Document Server

Judd, Anthony M

2014-01-01

An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

International Nuclear Information System (INIS)

2002-06-01

All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

Coolant mixing in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Hoehne, T; Grunwald, G

1998-10-01

The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Schatz, K.

1982-01-01

In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

Sodium coolant of fast reactors: Experience and problems

International Nuclear Information System (INIS)

Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

1997-01-01

In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

International Nuclear Information System (INIS)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs

Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

Energy Technology Data Exchange (ETDEWEB)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

International Nuclear Information System (INIS)

Kawai, H.

2001-01-01

JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report

International Nuclear Information System (INIS)

Fair, C.E.; Greer, A.O.

1983-03-01

This report presents a set of guidelines and criteria for improving main coolant pump shaft seal operational reliability. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies. Usage procedures/practices and operational environment influence on seal life and reliability from the most recent such survey are summarized. The shaft seal and its auxiliary supporting systems are discussed both from technical and operational related viewpoints

Transient performance of flow in PWR reactor circuits

International Nuclear Information System (INIS)

Hirdes, V.R.T.R.; Carajilescov, P.

1988-12-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

International Nuclear Information System (INIS)

Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

2003-01-01

Full text: The work that has been conducted so far justifies a possibility of constructing a reactor with a non-traditional coolant to develop radically new reactors and their cycles with perfect architecture. A solid coolant, for example, the carbon-based one, allows to design the primary circuit of nuclear reactor without excess pressure. Such coolant withstands temperatures up to ∼4000 deg. K without a collapse. The analysis of theory and experiments produced requirements to be met by a solid coolant used in the primary circuit of nuclear reactor. One of the most important requirements is the arrangements for a continuous and homogeneous gravity flow of the coolant through all core sections taking into account the dust caused by wear and some amount of fractured particles. Therefore, the idea is that the mass of particles should resemble a liquid to a certain extend. The particles should be sphere like with average diameter from 0.5 to 2.0 mm and nonsphericity rate not more than 10%. 'Angle of repose' of particles to the horizon can be utilised as a validity criterion of particles which should not exceed 25 deg. The heat transfer coefficient should be increased up to the practical maximum value. In 1996 - 1997 the system of experimental facilities were built in the Scientific and Research Institute 'Luch' to prove the possibility to reliably cool a nuclear reactor with a flow of solid particles and to obtain a minimum set of data for the conceptual design of such reactor with solid coolant. The facility allows the research of the flow stability, heat mass transfer in the core, lifetime wearing of particles of the solid coolant. In 1994-1999 5 batches of particles of different size were fabricated in accordance to different technologies. Four batches were graphite-based and one was aluminium oxide-based (Al 2 O 3 ). The purpose was to verify how the heat transfer coefficient was changing as the particle size varied. The average diameter of graphite particles

System and method for determining coolant level and flow velocity in a nuclear reactor

Science.gov (United States)

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

International Nuclear Information System (INIS)

Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

1999-01-01

During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

Reactor recirculation pump test loop

International Nuclear Information System (INIS)

Taka, Shusei; Kato, Hiroyuki

1979-01-01

A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

International Nuclear Information System (INIS)

Costa, J.R.

1978-12-01

An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

Organic coolant for ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

1991-04-01

ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

Process and kinetics of the fundamental radiation-electrochemical reactions in the primary coolant loop of nuclear reactors

International Nuclear Information System (INIS)

Kozomara-Maic, S.

1987-06-01

In spite of the rather broad title of this report, its major part is devoted to the corrosion problems at the RA reactor, i.e. causes and consequences of the reactor shutdown in 1979 and 1982. Some problems of reactor chemistry are pointed out because they are significant for future reactor operation. The final conclusion of this report is that corrosion processes in the primary coolant circuit of the nuclear reactor are specific and that radiation effects cannot be excluded when processes and reaction kinetics are investigated. Knowledge about the kinetics of all the chemical reactions occurring in the primary coolant loop are of crucial significance for safe and economical reactor operation [sr

Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

International Nuclear Information System (INIS)

Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

1996-01-01

The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

International Nuclear Information System (INIS)

Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

2001-01-01

In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

Use of a multi-attribute utility theory for evaluating the best coolant material in transmutation reactors

International Nuclear Information System (INIS)

Yu, Dong Han; Han, Suk Joong; Kim, Do Hyung; Park, Won Suk

1998-12-01

In order to develop and design a good transmutation system, it is necessary first to select the best available coolant material for a reactor coolant system. Choosing the best coolant material may not be easy since there are several criteria associated with thermal performance, safety problem, cost problem, neutronic aspects. etc. The best option should be chosen based on the maximization of our needs in this situation. It is a challenging task. Decision theory can be employed to solve this type of problem. This report presents the feasibility study for evaluating the best coolant material in transmutation reactors based on the multi=attribute utility theory. The main problem presented here is how to logically evaluate candidate coolant materials under multiple criteria such as thermal performance, safety problem, cost problem, cost problem, neutronic aspects, etc. Since the current problem involves multiple criteria or attributes, first of all, the multi-attribute utility theory (MAUT) such as SMART and AHP has been extensively reviewed. Then, many candidate coolant material for transmutation reactors have been identified. The next step is to construct a value tree that express to reflect the relative importance of the attributes for overall evaluation. Finally, given these assignments, the final goal were obtained by manipulating these ranks through the value tree. The proposed approach is intended to help people be rational and logical in making decisions such complex tasks. (author). 8 refs., 7 tabs., 22 figs

Fast breeder reactor electromagnetic pump

International Nuclear Information System (INIS)

Araseki, Hideo; Murakami, Takahiro

2008-01-01

Main pumps circulating sodium in the FBR type reactor have been mechanical types, not electromagnetic pumps. Electromagnetic pump of 1-2 m 3 /min has been used as an auxiliary pump. Large sized electromagnetic pumps such as several hundred m 3 /min have not been commercialized due to technical difficulties with electromagnetic instability and pressure pulsations. This article explained electromagnetic and fluid equations and magnetic Reynolds number related with electromagnetic pumps and numerical analysis of instability characteristics and pressure pulsations and then described applications of the results to FBR system. Magnetic Reynolds number must be chosen less than one with appropriate operating frequency and optimum slip of 0.2-0.4. (T. Tanaka)

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

International Nuclear Information System (INIS)

Curtis, Franklin G.; Ekici, Kivanc; Freels, James D.

2011-01-01

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

2016-01-15

A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

Monitoring for shaft cracks on reactor recirculation pumps

International Nuclear Information System (INIS)

Kowal, M.G.; O'Brien, J.T. Jr.

1989-01-01

The article discusses the vibration characteristics associated with a boiling water reactor (BWR) recirculation pump. It also describes the application of diagnostic techniques and shaft crack theory to an on-line diagnostic monitoring system for reactor recirculation pumps employed at Philadelphia Electric Company's Peach Bottom Atomic Power Station. Specific emphasis is placed on the unique monitoring techniques associated with these variable speed vertical pumps

Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

2012-01-01

The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net-power...

Method of determination of thermo-acoustic coolant instability boundaries in reactor core at NPPs with WWER

International Nuclear Information System (INIS)

Skalozubov, Volodymyr; Kolykhanov, Viktor; Kovryzhkin, Yuriy

2007-01-01

The regulatory body of Ukraine, the National Atomic Energy Company and the Scientific and Production Centre have led team-works concerned with previously unstudied factors or phenomena affecting reactor safety. As a result it is determined that the thermo-acoustic coolant instability conditions can appear in the core at definite operating WWER regimes. Considerable cyclic dynamic loads affect fuel claddings over thermo-acoustic pressure oscillations. These loads can result in inadmissible cassette design damage and containment damage. Taking into account calculation and experimental research authors submit a method of on-line assessment of WWER core state concerning thermo-acoustic coolant instability. According to this method, the thermo-acoustic coolant instability appearance conditions can be estimated using normal registered parameters (pressure, temperature, heat demand etc.). At operative modes, a WWER-1000 core is stable to tracheotomies oscillations, but reduction of coolant discharge through the core for some times can result in thermo-acoustic coolant instability. Thermo-acoustic instability appears at separate transitional modes concerned with reactor scram and unloading/loading at all power units. When thermo-acoustic instability begins in transitional modes, core elements are under influence of high-frequency coolant pressure pulsations for a long time (tens of hours)

Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

Energy Technology Data Exchange (ETDEWEB)

Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

2012-07-01

In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

Experience on vibration analysis of primary coolant pumps in Cirus

International Nuclear Information System (INIS)

Ullas, O.P.; Tilara, Manoj; Kharpate, A.V.

2002-01-01

Full text: 40 MW (thermal) CIRUS research reactor has been in operation for over four decades. During the major portion of its life almost all the major mechanical equipment operated continuously in a healthy condition. Since 1988 ageing related breakdown has been noticed in some of the critical components, PCW pumps being one of them. Vibration measurement and analysis is carried out on a routine basis as a part of conditioning monitoring programme. Ageing degradation of various components of the pump has been detected by such a performance monitoring programme. Conditioning monitoring has been found to be quite useful for scheduling of maintenance work on pumps

Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

International Nuclear Information System (INIS)

Liebig, E.

1977-01-01

A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

Transient performance of flow in circuits of PWR type reactors

International Nuclear Information System (INIS)

Hirdes, V.R.; Carajilescov, P.

1988-09-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Braun, H. E.; Dollard, W. J.; Tower, S. N.

1980-01-01

A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

Phenomena occurring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1989-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

International Nuclear Information System (INIS)

Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

1984-07-01

The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

International Nuclear Information System (INIS)

Sharma, V.K.; Surendar, C.; Bapat, C.N.

1994-01-01

The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

International Nuclear Information System (INIS)

Victor Hugo Sanchez Espinoza

2005-01-01

Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during

Development of sputter ion pump based SG leak detection system for Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Babu, B.; Sureshkumar, K.V.; Srinivasan, G.

2013-01-01

Highlights: ► Development and commissioning of SG leak detection system for FBTR. ► Development of Robust method of using sputter ion pump based system. ► Modifications for improving reliability and availability. ► On line injection of hydrogen in sodium during reactor operation. ► Triplication of the SG leak detection system. - Abstract: The Fast Breeder Test Reactor (FBTR) is a 40 MWt, loop type sodium cooled fast reactor built at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam as a fore-runner to the second stage of Indian nuclear power programme. The reactor design is based on the French reactor Rapsodie with several modifications which include the provision of a steam-water circuit and turbo-generator. FBTR uses sodium as the coolant in the main heat transport medium to transfer heat from the reactor core to the feed water in the tertiary loop for producing superheated steam, which drives the turbo-generator. Sodium and water flow in shell and tube side respectively, separated by thin-walls of the ferritic steel tubes of the once-through steam generator (SG). Material defects in these tubes can lead to leakage of water into sodium, resulting in sodium water reactions leading to undesirable consequences. Early detection of water or steam leaks into sodium in the steam generator units of liquid metal fast breeder reactors (LMFBR) is an important requirement from safety and economic considerations. The SG leak in FBTR is detected by Sputter Ion Pump (SIP) based Steam Generator Leak Detection (SGLD) system and Thermal Conductivity Detector (TCD) based Hydrogen in Argon Detection (HAD) system. Many modifications were carried out in the SGLD system for the reactor operation to improve the reliability and availability. This paper details the development and the acquired experience of SIP based SGLD system instrumentation for real time hydrogen detection in sodium for FBTR.

Loss of coolant accident at boiling water reactors

International Nuclear Information System (INIS)

Ramirez G, R.

1975-01-01

A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

Determination of temperature distributions in fast reactor core coolants