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Sample records for safety rods

  1. Safety rod driving device

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Kurosaki, Akira.

    1988-01-01

    Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)

  2. Cadmium safety rod thermal tests

    International Nuclear Information System (INIS)

    Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

    1992-01-01

    Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

  3. Temperature actuated automatic safety rod release

    Science.gov (United States)

    Hutter, E.; Pardini, J.A.; Walker, D.E.

    1984-03-13

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  4. Temperature actuated automatic safety rod release

    International Nuclear Information System (INIS)

    Hutter, E.; Pardini, J.A.; Walker, D.E.

    1987-01-01

    This patent describes a nuclear reactor having a core, a safety rod for downward insertion into and upward withdrawal from the core, a drive shaft for supporting and operating the safety rod, and drive means connected to the drive shaft for operating the shaft. An apparatus is described for releasably supporting the safety rod, the apparatus comprising an upper adapter adapted to be affixed to the upper end of the safety rod, the upper adapter having a retention means, a lower portion on the drive shaft and having a hollow interior for housing the upper adapter, a bimetallic means supported within the hollow interior of the lower portion and having at least one ledge which engages the retention means to support the upper adapter, the bimetallic means being a substantially cylindrical bimetallic member for receiving the upper adapter in a generally coaxial relation, the substantially cylindrical bimetallic member comprising an inner layer and an outer layer, and the inner layer having a greater coefficient of thermal expansion than the outer layer

  5. Reactivity shimming in FFTF with safety rods

    International Nuclear Information System (INIS)

    Bennett, C.L.

    1986-01-01

    The Fast Flux Test Facility (FFTF) reactor demonstrated operation at a 65% annual capacity factor in 1984 and a 70% annual capacity factor in 1985. This was achieved in part by extending the cycle lengths beyond the originally designed 100-day cycles. The advent of extended cycle lengths at the FFTF reactor led to the development of a contingency plan to shim some of the installed excess reactivity at the beginning of the cycle with the primary control rod bank (safety rods) in order to maintain the minimum required shutdown margin in the secondary shutdown system. This paper describes both the implementation of this plan in terms of the licensing aspects and the actual use of primary shim during cycles 5 and 6 operation at FFTF

  6. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

    The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

  7. Measurement and analysis of CEFR safety and shim rod worth

    International Nuclear Information System (INIS)

    Chen Yiyu; Yang Yong; Gang Zhi; Xu Li; Yang Xiaoyan; Zhou Keyuan; Hu Dingsheng

    2013-01-01

    The reactivity worth of safety rods and shim rods in critical phase and operating phase was calculated respectively using Monte Carlo program in this paper. In addition, the reactivity worth of safety rods and shim rods was measured by the rod drop-off method and period method. The experimental results are in good agreement with the calculated values with less than 5% error. It illustrates the high calculation precision of Monte Carlo program, which provides a practical reference for subsequent application of Monte Carlo program in future demonstration fast reactors. (authors)

  8. Substitute safety rods: Physics design and NTG calibration

    International Nuclear Information System (INIS)

    Baumann, N.P.

    1991-07-01

    Under certain assumed accident conditions, an SRS reactor may loose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the safety rod. Tests have shown that the current cadmium safety rod, which contains aluminum as well as cadmium, can fail at temperatures only slightly in excess of 500 deg C. Computations indicate that such temperatures can be reached with operating powers well below the 50% power limit now imposed by other accident scenarios. Safety rod melting would thus establish a new lower operating limit. A substitute safety rod that could tolerate much higher temperatures would eliminate this limit. This memorandum details the physics characteristics of a suitable replacement rod. 7 refs

  9. Motion simulation of hydraulic driven safety rod using FSI method

    International Nuclear Information System (INIS)

    Jung, Jaeho; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2013-01-01

    Hydraulic driven safety rod which is one of them is being developed by Division for Reactor Mechanical Engineering, KAERI. In this paper the motion of this rod is simulated by fluid structure interaction (FSI) method before manufacturing for design verification and pump sizing. A newly designed hydraulic driven safety rod which is one of reactivity control mechanism is simulated using FSI method for design verification and pump sizing. The simulation is done in CFD domain with UDF. The pressure drop is changed slightly by flow rates. It means that the pressure drop is mainly determined by weight of moving part. The simulated velocity of piston is linearly proportional to flow rates so the pump can be sized easily according to the rising and drop time requirement of the safety rod using the simulation results

  10. Additional information for impact response of the restart safety rods

    International Nuclear Information System (INIS)

    Yau, W.W.F.

    1991-01-01

    WSRC-RP-91-677 studied the structural response of the safety rods under the conditions of brake failure and accidental release. It was concluded that the maximum impact loading to the safety rod is 6020 pounds based on conservative considerations that energy dissipation attributable to fluid resistance and reactor superstructure flexibility. The staffers of the Defense Nuclear Facility Safety Board reviewed the results and inquired about the extent of conservatism. By request of the RESTART team, I reassessed the impact force due to these conservative assumptions. This memorandum reports these assessments

  11. Substitute safety rods: Physics of operation and irradiation

    International Nuclear Information System (INIS)

    Baumann, N.P.

    1991-01-01

    Under certain assumed accidents, an SRS reactor may lose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the currently used cadmium safety rod. A substitute safety rod consisting solely of sintered B 4 C and stainless steel has been designed which is capable of withstanding much higher temperatures. This memorandum provides the physics basis for the adequacy of the rod for reactor shutdown and provides a set of criteria for acceptance in the NTG tests. This memorandum provides physics data for other aspects of operation. These include: Heat production and helium production, along with related phenomena, resulting from inadvertent irradiation at power. Gamma heat input under drained tank conditions. An equivalent rod design suitable for charge design and safety analyses. Degradation under normal operation. Thermal flux ripple in adjacent fuel due to axial striping of alternate B 4 C and steel pellets. Possible effect on safety analyses. Safety rod withdrawal during reactor startup

  12. Ameliorative design for CARR safety rod drive mechanism

    International Nuclear Information System (INIS)

    Zhu Xuewei; Luo Zhong; Zhen Jianxiao; Wang Yulin

    2014-01-01

    The problem of safety rod accident dropped during C commissioning phase for China Advanced Research Reactor (CARR) was analyzed, and the reason was that the solenoid valve in safety rod drive mechanism (SRDM) driven loop was breakdown because of long-playing work. To solve this safe hidden trouble, SRDM was redesigned, and a new type of 'hydro lifting-hydro and electromagnetic holding' SRDM was presented, using Ansoft Maxwell to make a finite element analysis on new SRDM, working out electromagnetic field distribution and electromagnetic force of new SRDM. The results show that the value of electromagnetic force produced by electromagnetic force holding unit reaches 2.12 times about the weight of safety rod drive line, and it has some margins. (authors)

  13. Seismic analysis of control and safety rod drive mechanism

    International Nuclear Information System (INIS)

    Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

    2003-01-01

    Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)

  14. Analysis of addition of the safety rods at RSG-GAS core

    International Nuclear Information System (INIS)

    S, Tukiran; S, Tagor Malem; K, Iman

    2002-01-01

    The silicide fuel loading of the RSG-GAS core is planned to increase from 250 gU to 300 gU. Increasing of fuel loading will prolong the operation cycle length from 25 days to 32,5 days, but ability of reactivity compensation by control rods system decreased because the reactivity shut-down margin is available only 1,03 %, expectation is 2.2 %. One of solutions is added two safety control rods in B-3 and G-10 positions the aim of installing two safety rods (BKP) in RSG-GAS core is to increase core safety margin. So before using the safety control rods in the RSG-GAS core, it is necessary to know its performance, one of the tests showing its performance is to measure the reactivity of the safety control rods. Measurement of safety control rods were done to know each reactivity worth of safety control rods at middle cycle so that the safety rod be used in the RSG-GAS core. Measurement done by using calibration control rods with couple compensation method which always using in the RSG-GAS core to measure the existing control rods. The results of measurement showed that two safety rods (BKP01 and BKP02) have reactivity worth of 93.5 cent and 87.5 cent, respectively. the total reactivity worth of safety control rods is 1.38%. So the two safety rods can be used to increase safety margin of the RSG-GAS core if the fuel is exchanged to 300 gU of loading

  15. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  16. Measurement of safety-rod effectiveness of the zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Popovic, D; Takac, S; Markovic, H; Martinc, R; Radmanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    The reactivity effectiveness of the two safety rods displaced eccentrically along diameter of a cylindrical D{sub 2}O moderated reactor was measured and compared with the theoretical calculations. The results show that the simplified calculations of one rod effectiveness are quite satisfactory but the theoretical evaluation of the interference effect of the two rods are not sufficiently reliable. (author)

  17. Tensile and burst tests in support of the cadmium safety rod failure evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.

    1992-02-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

  18. Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

    1992-01-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

  19. Safety coupling for a control rod of a nuclear reactor

    International Nuclear Information System (INIS)

    Mindnich, F.R.; Friedrichs, H.; Schoettle, J.

    1978-01-01

    A coupling is presented between a control rod and the drive shafft arranged below. The construction of this coupling is designed in such a way that the usual sealing maesures against the escape of coolant are reduced. (TK) [de

  20. Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

    International Nuclear Information System (INIS)

    Shaw, K.W.

    1993-11-01

    The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

  1. Development of embedded Control System for Control and Safety Rod Drive Mechanisms (CSRDMs) of PFBR

    International Nuclear Information System (INIS)

    Kameswari, K.; Palanisami, K.; Thirugnana Murthy, D.; Murali, N.; Satyamurty, S.A.V.

    2013-01-01

    Prototype Fast Breeder Reactor (PFBR), a 500 MWe, Sodium cooled, fast breeder reactor is nearing completion at Kalpakkam, Tamil Nadu. PFBR has two independent, fast acting and diverse shutdown systems, one with nine Control and Safety Rods (CSRs) and another with three Diverse Safety Rods (DSRs), with independent driving mechanisms called CSRDMs and DSRDMs respectively. This paper deals with the development of Real Time Computer based Control system for controlling nine CSRDMs with model based software development environment - SCADE (Safety Critical Application Development Environment). (author)

  2. Radiological safety aspects during handling of adjuster rod (cobalt-60) samples at Cirus

    International Nuclear Information System (INIS)

    Yadav, R.K.B.; Prasad, S.K.; Khuspe, R.R.; Babu, K.S.; Kamble, M.K.; Deshpande, S.B.; Srivastava, Alok; Ramesh, N.; Sharma, R.C.

    2008-01-01

    Cobalt -60 produced from the irradiation of the 59 Co capsules in adjuster rod of Cirus is used in many medical and industrial applications such as radiation therapy, sterilization of agricultural products, sterilization of medical equipment, industrial gamma exposure devices, nucleonic gauging sources etc. Two adjuster rods used for adjusting small reactivity loads (3.5 mk at 40 MW) in reactor contain thirty 59 Co slugs each. The 59 Co slugs are in the form of small capsules. These slugs were irradiated for nearly 3 years of full power operation. Since the amount of activity handled in the adjuster rod is very high (35 - 40 kCi), adequate radiological safety coverage is to be provided during the transfer of adjuster rod from the reactor pile to Tray Rod Facility (TRF) and during remote handling and loading of adjuster rod samples in shipping flask at TRF. Radiation fields observed are 0.4 - 3.0 Gy/h. In Cirus, falling of Adjuster rod from vertical flask at top of pile could result in declaration of evacuation emergency due to potential hazards associated with the high radiation field (∼26 Gy/h at 1m) of 60 Co slugs. Entire process of remote handling is viewed through lead glass and movement of samples is carried out with the help of a pair of master slave manipulator (MSM) arms. TRF with full samples give a radiation background of 0.2 - 20 mGy/h, requiring control of movement of personnel in the area. This paper explains in detail about the health physics aspects during handling of adjuster rod from reactor pile to TRF and back to pile, unloading and loading of radioisotope samples, operational radiation protection experience gained and collective dose consumed for the above transfer and loading job. The collective dose consumed in the handling of two adjuster rods is 8.7 per-mSv. (author)

  3. Design of diverse safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Vijayashree, R.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    In Prototype Fast Breeder Reactor (PFBR), there are two types of absorber rods for control and shutdown of the reactor in the event of any abnormal event. They are: (i) Control and Safety Rod (CSR) and (ii) the Diverse Safety Rod (DSR). Of these, the former (CSR) caters to the control function of the reactor during normal operating conditions and to the shutdown during abnormal situations. The DSR, on the other hand is meant essentially for the reactor shutdown to take care of any abnormal transient. It is rather important to note that functionally the DSR is independent of CSR in the sense, that it can bring the reactor to a cold shutdown state and maintain it even under the hypothetical condition of the failure of CSR. From the design point of view, this stipulates a failure probability of less than 10 -4 per demand. The DSR is normally parked above the core by the Diverse Safety Rod Drive Mechanism (DSRDM). On receiving a scram signal it gets released from the holding electromagnet and falls under the gravity into the core. Diverse features are incorporated both in the absorber rods and in the drive mechanisms to avoid common mode failures. This paper discusses the salient features of DSR and DSRDM. A brief account of detailed design, analysis and development of two important subassemblies viz. electromagnet and sodium dash pot is also presented. In addition, a brief comparison between CSR and DSR including their drive mechanisms is also provided. (author)

  4. Experiments with preirradiated fuel rods in the Nuclear Safety Research Reactor

    International Nuclear Information System (INIS)

    Horiki, O.; Kobayashi, S.; Takariko, I.; Ishijima, K.

    1992-01-01

    In the Nuclear Safety Research Reactor (NSRR) owned and operated by Japan Atomic Energy Research Institute (JAERI), extensive experimental studies on the fuel behavior under reactivity initiated accident (RIA) conditions have been continued since the start of the test program in 1975. Accumulated experimental data were used as the fundamental data base of the Japanese safety evaluation guideline for reactivity initiated events in light water cooled nuclear power plants established by the nuclear safety commission in 1984. All of the data used to establish the guideline were, however, limited to those derived from the tests with fresh fuel rods as test samples because of the lack of experimental facility to handle highly radioactive materials.The guideline, therefore, introduces the peak fuel enthalpy of 85 cal/g which was adopted from the SPERT-CDC data as a provisional failure threshold of preirradiated fuel rod and, says that this value should be revised based on the NSRR experiments in the future. According to the above requirement, new NSRR experimental program with the preirradiated fuel rods as test samples was started in 1989. Test fuel rods are prepared by refabrication of the long-sized fuel rods preirradiated in commercial PWRs and BWRs into short segments and by preirradiation of short-sized test fuel rods in the Japan Material Testing Reactor(JMTR). For the tests with preirradiated fuel rods as test samples, the special experimental capsules, the automatic instrumentation fitting device, the automatic capsule assembling device and the capsule loading device were newly developed. In addition, the existing hot cave was modified to mount the capsule assembling device and the other inspection tools and, a new small iron cell was established adjacent to the cave to store the instrumentation fitting device. (author)

  5. Design of control and safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    Control and Safety Rod (CSR) is one of the two types of absorber rods in shutdown systems of PFBR. Control and Safety Rod Drive Mechanism (CSRDM) actuates CSR to have vertical translatory motion in reactor core. The dual responsibilities entrusted on CSR to control reactor power during normal operating condition and to shutdown the reactor by scram action during abnormal condition, necessitate highly reliable design, analysis, testing and surveillance of CSR and CSRDM. The paper discusses on the salient features of CSR and CSRDM and design and analysis of individual sub-assemblies, viz., gripper, scram-release electromagnet, hydraulic dash pot, seals. Also it discusses on the developmental activities proposed and surveillance test requirements. (author)

  6. A new physics design of control safety rods for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

    2008-01-01

    The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

  7. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  8. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  9. Probabilistic safety analysis for control rod drive system of ET-RR-1

    International Nuclear Information System (INIS)

    Nasr, M.; Nasser, O.

    1988-01-01

    The International Atomic Energy Agency (IAEA) co-ordinated a Research programme on Probabilistic Safety Analysis (PSA) for research reactors; with the participation of several countries. In the framework of this project (Project Int. 9/063) the Egyptian Atomic Energy Authority decided to perform a PSA study on the ET-RR-1 (Egypt Thermal Research Reactor). The study is conducted in collaboration between the nuclear regulatory and safety centre (NRSC) and the reactor department of the nuclear research centre at Inchass. The present work is a part of the PSA study on ET-RR- it is concerning a probabilistic safety analysis of the control rod drive mechanism

  10. Cadmium control/safety rod disposal at the Savannah River Site

    International Nuclear Information System (INIS)

    McInnis, S.H.

    1995-01-01

    Four heavy-water-moderated reactors at the Savannah River Site will undergo the removal of 862 activated cadmium control/safety rods. Although these reactors are 40 years old, they offer 4 basic advantages for decommissioning: the equipment is still in some sort of operable state; the reactor is blow the floor in a large process room, allowing access; Control/safety rods can be handled remotely by existing equipment; a radiologically shielded removal path exists. Drawbacks include the following: age of reactors; improvements in technology have caused incompatibility problems; more strigent standards; compliance with environmental regulations. This article details how the removal was carried out and the current status of the project, keeping in mind the above considerations

  11. New control and safety rod unit for the training reactor of the Dresden Technical University

    International Nuclear Information System (INIS)

    Adam, E.; Schab, J.; Knorr, J.

    1983-01-01

    The extension of the experimental training of students at the training reactor AKR of the Dresden Technical University requires the reconstruction of the reactor with a new control and safety rod unit. The specific conditions at the AKR led to a new variant. Results of preliminary experiments, design and mode of operation of the first unit as well as hitherto gained operation experiences are presented. (author)

  12. Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Thanigaiyarasu, G.; Chellapandi, P.

    2014-01-01

    Highlights: • Mathematical modelling of dynamic behaviour of safety rod during scram action in fast reactor. • Effects of hydraulics, structural interaction and geometry on drop time of safety rod are understood. • Using simplified model, drop time can be assessed replacing detailed CFD analysis. • Sensitivities of the related parameters on drop time are understood. • Experimental validation qualifies the modelling and computer software developed. - Abstract: Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500 MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated

  13. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  14. Safety of 5 MW district heating reactor (DHR) and hydraulic dynamic pressure drive control rods

    International Nuclear Information System (INIS)

    Wu Yuanqiang; Wang Dazhong

    1991-11-01

    The principles and movement characteristic of the hydraulic dynamic pressure drive for control rods in 5 MW district heating reactor are described with stress on analysis of its effects on reactor safety features. The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR. The drive cylinder is driven by dynamic pressure. In the new drive system, the reactor coolant (water) used as actuating medium is pressed by pump, then injected into a step cylinder which is set in the reactor core. The cylinder will move step by step by controlling flow, then the cylinder drives the neutron absorber and controls nuclear reaction. The drive is characterized by simplicity in structure, high reliability, inherent safety, reduction in reactor height, economy, etc

  15. Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Selby, G.P.

    1978-09-01

    If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

  16. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    Energy Technology Data Exchange (ETDEWEB)

    Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, Zanjan, P.O. Box 451-313 (Iran, Islamic Republic of); Sadat Kiai, S.M., E-mail: sadatkiai@yahoo.co [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I., P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, Tehran (Iran, Islamic Republic of)

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity (rho), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  17. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO 2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  18. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    1997-01-01

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  19. Safety analysis results for the control rod banks withdrawal event at a full power of the SMART-P

    International Nuclear Information System (INIS)

    Yang, S. H.; Chung, Y. J.; Kim, H. C.; Zee, S. Q.

    2005-01-01

    For the validation of the 330 MWt SMART (System-integrated Modular Advanced ReacTor), a detailed design for the SMART-P has been accomplished by KAERI. In the SMART-P design similar to the SMART design, the soluble boron free design is adapted. This concept results in a larger reactivity worth of the control rod bank compared to that of the commercial pressurized water reactor. Moreover, in the SMART-P design, the control rod banks are fairly well inserted into the core, even at a full power condition. Therefore, accidents related to the reactivity anomalies have been evaluated as crucial events when compared to the other initiating events. In this paper, safety analysis for the control rod banks withdrawal event at a full power of the SMART-P has been accomplished by considering various initial conditions, different withdrawal times of the control rod banks and the reactivity feedback. To perform the safety analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code for a system response and SSF-1 correlation for a CHFR (Critical Heat Flux Ratio) have been used

  20. Safety and efficacy of Implanon, a single-rod implantable contraceptive containing etonogestrel.

    Science.gov (United States)

    Funk, Sidney; Miller, Michael M; Mishell, Daniel R; Archer, David F; Poindexter, Alfred; Schmidt, Juergen; Zampaglione, Edio

    2005-05-01

    The safety and efficacy of a single-rod implantable contraceptive containing etonogestrel (Implanontrade mark) were investigated in a multicenter clinical trial. Sexually active American women (N=330) with apparently normal menstrual cycles used the implant for up to 2 years. All subjects recorded bleeding and/or spotting daily in a diary. Safety was assessed through adverse experiences (AEs), laboratory tests and physical and gynecologic examinations. Total exposure was 474 woman-years (6186 cycles), and 68% of subjects had at least 1 year of exposure. No pregnancies occurred. The most common bleeding pattern observed throughout the study was infrequent bleeding, defined as less than three episodes of bleeding in a reference period (excluding amenorrhea). The least common pattern was frequent bleeding, defined as more than five episodes of bleeding in a reference period. Infrequent, prolonged and frequent bleeding patterns were most common early in the study and declined thereafter. During the 3-month Reference Periods 2-8 (Months 4-24), the incidence of amenorrhea ranged from 14% to 20%. Forty-three subjects (13%) withdrew from the study because of bleeding pattern changes and 76 subjects (23%) discontinued because of other AEs. Other common AEs leading to discontinuation, besides bleeding irregularities, were emotional lability (6.1%), weight increase (3.3%), depression (2.4%) and acne (1.5%). Use of Implanon (etonogestrel subdermal implant, referred to herein as ENG implant) for up to 2 years had no clinically significant effects on laboratory parameters, physical and pelvic examinations, vital signs or body mass index. The average length of time required for ENG implant insertion and that for removal were 0.5 and 3.5 min, respectively, and all the procedures were uncomplicated. The return to normal menstrual cycles and fertility was rapid after removal. Implanon is a safe, highly effective and rapidly reversible new method of contraception.

  1. Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

    International Nuclear Information System (INIS)

    Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

    1992-01-01

    The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

  2. Safety techniques of lightning rod and static electricity in oil tanks and oil trucks

    International Nuclear Information System (INIS)

    Ilievska, Tatjana

    1999-01-01

    In this article the ways and examples of lightning rod installation of small tanks for storage of both oil and oil derivates used by petrol stations are presented (an example of some petrol stations in the wider region in Bitola is given ). Also, a lightning rod protection of big tanks and terminals as well as protection of static electricity of tank trucks during transportation of fuel is represented. Special review is given to the protection of static electricity during transforming (decanting) of the fuel. (Author)

  3. Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

    2009-02-15

    A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.

  4. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  5. On the Rod Drop technique in integral reactivity measures in control banks and reactor safety

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo

    2013-01-01

    This work presents a study on the effect of shading in neutron detectors, when used in measures of reactivity with the rod drop technique. Shading can be understood as a change in the efficiency of the detectors, when it is given in detected neutrons fission occurred in the reactor, more evident in the detectors closest to the bank being inserted. The method of analysis was based on simulations of reactor IPEN/MB-01, using the code CITATION and MCNP program. In both cases, the results were static, showing Neutronic flows in only two situations: before insertion of the control rod and after insertion. The measure of reactivity in this case was achieved using the expression derived from the source jerk technique. In addition to theoretical study, data from a rod drop experiment conducted in the reactor IPEN/MB-01 were also used. In this case, the reactivity was obtained using inverse kinetic method, since experimental data were set of values that vary with time. In all cases, correction factors for the shadowing effect have been proposed. (author)

  6. Cooling of safety rods in the Savannah River K Reactor during the gamma heating phase of a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Unal, C.; Motley, F.E.; Rodriguez, S.B.

    1992-01-01

    This paper documents the heat-transfer analysis for the safety rod placed in a perforated guide tube during the gamma heating phase of a large-break loss of coolant accident in Savannah River K-reactor. The cooling mechanisms are natural convection to air and radiation to the surrounding structures. The limiting component is the guide tube. The guide tube is shown to remain coolable below its thermal limit for the anticipated reactor powers unless it is contacted by the hotter safety rod. Sample calculations are performed for various contact scenarios, and the results are reported within the paper. The results indicate that the most limiting contact scenario results when the safety rod heats up to its maximum temperature while remaining concentric in the guide tube and then contacts the guide tube. The worse contact location appears to be in line with the slugs-cladding contact and in between the rows of holes in the guide tube

  7. Report of a consultants meeting on control rod insertion reliability for WWER-1000 nuclear power plants. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1995-09-01

    Starting from 1992, an increased drop time of control rods exceeding the design limit of four seconds has been observed in most of the operating WWER-1000 reactors in Russia and in the Ukraine. In some cases a dropped control rod became stuck in an intermediate position near the bottom of the core. In October 1994, a similar control rod problem was also observed at Unit 6 of the Kozloduy NPP. The issue of control rod insertion reliability was considered at a consultants' meeting on ''Core Control and Protection Strategy of WWER-1000 Reactors'' in April 1994. A consultants' meeting specifically focused on ''Control Rod Insertion Reliability'' was convened in Vienna in February 1995 attended by 15 international experts. The objectives of this meeting were: The exchange of international experience on problems and solutions related to anomalous control rod insertion; judgement of the safety concern of this issue for WWER-1000 reactors based on safety analyses; consideration of regulatory requirements and interim measures to continue operation in short term including modifications implemented or planned; and, status of root cause analyses and pending problems. The technical discussions were held in plenary sessions and in three working groups devoted to specific aspects of the issue. Refs, figs, tabs

  8. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

    Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

  9. Control rod drives

    International Nuclear Information System (INIS)

    Futatsugi, Masao.

    1980-01-01

    Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)

  10. Control rods

    International Nuclear Information System (INIS)

    Maruyama, Hiromi.

    1984-01-01

    Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

  11. Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

    International Nuclear Information System (INIS)

    Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

    1979-01-01

    The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

  12. Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

  13. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

    1987-01-01

    Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

  14. Control rod drives

    International Nuclear Information System (INIS)

    Asano, Hiromitsu.

    1979-01-01

    Purpose: To drive control rods at an optimum safety speed corresponding to the reactor core output. Constitution: The reactor power is detected by a neutron detector and the output signal is applied to a process computer. The process computer issues a signal representing the reactor core output, which is converted through a function generator into a signal representing the safety speed of control rods. The converted signal is further supplied to a V/F converter and converted into a pulse signal. The pulse signal is inputted to a step motor driving circuit, which actuates a step motor to operate the control rods always at a safety speed corresponding to the reactor core power. (Furukawa, Y.)

  15. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

    This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

  16. Control rods

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1979-01-01

    Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

  17. Water rod

    International Nuclear Information System (INIS)

    Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

    1992-01-01

    In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

  18. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu; Kawamura, Atsuo.

    1979-01-01

    Purpose: To reduce pellet-clad mechanical interactions, as well as improve the fuel safety. Constitution: In the rod drive of a bwr type reactor, an electric motor operated upon intermittent input such as of pulse signals is connected to a control rod. A resolver for converting the rotational angle of the motor to electric signals is connected to the rotational shaft of the motor and the phase difference between the output signal from the resolver and a reference signal is adapted to detect by a comparator. Based on the detection result, the controller is actuated to control a motor for control rod drive so that fine control for the movement of the control rod is made possible. This can reduce the moving distance of the control rod, decrease the thermal stress applied to the control rod and decrease the pellet clad mechanical interaction failures due to thermal expansion between the cladding tube and the pellets caused by abrupt changes in the generated power. (Furukawa, Y.)

  19. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

    1988-01-01

    Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

  20. Measuring device for control rod driving time

    International Nuclear Information System (INIS)

    Tanaka, Kazuhiko; Hanabusa, Masatoshi.

    1993-01-01

    The present invention concerns a measuring device for control driving time having a function capable of measuring a selected control rod driving time and measuring an entire control rod driving time simultaneously. A calculation means and a store means for the selected rod control rod driving time, and a calculation means and a store means for the entire control rod driving time are disposed individually. Each of them measures the driving time and stores the data independent of each other based on a selected control rod insert ion signal and an entire control rod insertion signal. Even if insertion of selected and entire control rods overlaps, each of the control rod driving times can be measured reliably to provide an advantageous effect capable of more accurately conducting safety evaluation for the nuclear reactor based on the result of the measurement. (N.H.)

  1. Sucker rod motor

    Energy Technology Data Exchange (ETDEWEB)

    Radzalov, N N; Radzhabov, N A

    1983-01-01

    The motor consists of rollers mounted on the wellmouth and connected by a flexible rink. Reciprocating mechanism is in the form of a horizontal non-mobile single-side operation cylinder, inside which a plunger and rod are mounted. The working housing of the hydrocylinder is connected to a gas-hydr aulic batter, and when running is connected via plunger to the high pressure source; running in reverse it is connected with a safety valve and automatic control unit. The unit is equipped with a reducer and a mechanical transformer consisting of screw and nut, and which is shutoff with a single-side lining. The plunger rod consists of an auger-like unit. The high pressure source is provided by the injection line of the sucker rod that has been equipped with a reverse valve.

  2. Results of examinations of safety experimental rods of trap-like type irradiated in the BN-600 reactor

    International Nuclear Information System (INIS)

    Tarasikov, V.P.; Voznesenskij, R.M.; Rudenko, V.A.

    2001-01-01

    The results of post-irradiation examination are reported for three trap-like scram rods having been in operation in BN-600 reactor for 311, 331 and 348 EFPD (effective full power days). Compacted boron carbide enriched with 80 at.% 10 B is used as an absorber, zirconium hydride serves as a moderator; cladding are fabricated from steel 06Kh16N15M3B. The results obtained show that two zones are formed in the absorber material which are different in fracture mode and positioned at different distances from the moderator. Radiation damages of steel cladding are noted to be arranged nonuniformly through the height of the rod. The cladding-absorber interaction manifests itself in various ways, this is associated with various absorber burnups. The area of cladding-zirconium hydride interaction constitutes ∼ 30-80 μm. The swelling of the moderator is 4-5% and does not result in a loss of a cladding-moderator clearance [ru

  3. CONTROL ROD

    Science.gov (United States)

    Walker, D.E.; Matras, S.

    1963-04-30

    This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

  4. Rodding Surgery

    Science.gov (United States)

    ... Physical activity prior to surgery,  Length of the operation; anesthesia issues,  Reason for the choice of rod,  Time in the hospital,  Length of recovery time at home,  Pain management including control of muscle spasms,  The rehabilitation plan. ...

  5. Control rods

    International Nuclear Information System (INIS)

    Koga, Isao; Masuoka, Ryuzo.

    1979-01-01

    Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

  6. Duke Power Company's control rod wear program

    International Nuclear Information System (INIS)

    Culp, D.C.; Kitlan, M.S. Jr.

    1990-01-01

    Recent examinations performed at several foreign and domestic pressurized water reactors have identified significant control rod cladding wear, leading to the conclusion that previously believed control rod lifetimes are not attainable. To monitor control rod performance and reduce safety concerns associated with wear, Duke Power Company has developed a comprehensive control rod wear program for Ag-In-Cd and boron carbide (B 4 C) rods at the McGuire and Catawba nuclear stations. Duke Power currently uses the Westinghouse 17 x 17 Ag-In-Cd control rod design at McGuire Unit 1 and the Westinghouse 17 x 17 hybrid B 4 C control rod design with a Ag-In-Cd tip at McGuire Unit 2 and Catawba Units 1 and 2. The designs are similar, with the exception of the absorber material and clad thickness. There are 53 control rods per unit

  7. Control rod

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Inoue, Kotaro.

    1979-01-01

    Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

  8. Control rod

    International Nuclear Information System (INIS)

    Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

    1994-01-01

    Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

  9. The calculated research of a fuel rod thermo technical safety at various laws of energy liberation and temperature changes

    International Nuclear Information System (INIS)

    Azarov, S.I.

    1995-01-01

    Different questions of the thermo technical safety and efficiency at various thermal loads in steady-state and transitional operating conditions are discussed. Resources up to heat exchange crisis at impulse, step and harmonical action when changing of coolant temperature and energy liberation in fuel are shown

  10. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Futatsugi, Masao; Goto, Mikihiko.

    1976-01-01

    Purpose: To provide a control rod drive mechanism using water as an operating source, which prevents a phenomenon for forming two-layers of water in the neighbourhood of a return nozzle in a reactor to limit formation of excessive thermal stress to improve a safety. Constitution: In the control rod drive mechanism of the present invention, a heating device is installed in the neighbourhood of a pressure container for a reactor. This heating device is provided to heat return water in the reactor to a level equal to the temperature of reactor water thereby preventing a phenomenon for forming two-layers of water in the reactor. This limits formation of thermal stress in the return nozzle in the reactor. Accordingly, it is possible to minimize damages in the return nozzle portion and yet a possibility of failure in reactor water. (Kawakami, Y.)

  11. Sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Rylov, B M; Kostur, I N; Shcheigiy, B I; Sukhanov, V S

    1983-01-01

    As an addendum to A.s. USSR patent No 769087, this particular sucker rod utilizes a differential piston spring that has been attached outside the body of the auxiliary pump. The pump cylinder is attached to the intake line of the main pump. The lower part of the auxiliary pump is equipped with vertical slits, while the differential piston is equipped with a perforated pusher and support under the spring; it can also be shifted as necessary with respect to the vertical slits.

  12. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  13. WWER safety analysis in case of simultaneous positive reactivity introduction by control rods withdrawal and pure condensate injection

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiienko, I.; Khalimonchuk, V.

    2011-01-01

    Results of calculation assessment of positive reactivity insertion rate are presented under condition of simultaneous effect on reactivity of two provided by WWER project reactivity operation systems - extract of regulation group and decrease of boron acid concentration. Showed that maximal positive reactivity insertion rate is achieved at hot zero power and it's too less than limit value 0.07β eff /sec defined by normative document 'Nuclear safety regulation of NPP with pressurized water reactors'. For given reasons concluded that necessity of application of the extended protection PZ-2 actuation time is absent on interval of pure condensate transportation for full exclusion of possibility of positive reactivity insertion simultaneously by two different operation systems. (Authors)

  14. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  15. Control rod drives

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1984-01-01

    Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

  16. Control rod assembly

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

  17. Control rod displacement

    International Nuclear Information System (INIS)

    Nakazato, S.

    1987-01-01

    This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

  18. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  19. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

    Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

    1990-01-01

    Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

  20. Detection device for control rod interference

    International Nuclear Information System (INIS)

    Saito, Noboru.

    1984-01-01

    Purpose: To enable to detect the mechanical interference or friction between a control rod and a channel box automatically, simply and rapidly. Constitution: A signal from a gate circuit and a signal from a comparison mechanism are inputted into an AND circuit if a control rod has not been displaced by a predetermined distance within a prescribed time Δt after the output of an insertion or withdrawal signal for the control rod, by which a control-rod-interference signal is outputted from the AND circuit. Accordingly, the interference between the control rod and the channel box can be detected automatically, easily and rapidly. Furthermore, by properly adjusting the prescribed time Δt set by the gate circuit, the degree of the interference can also be detected, whereby the safety and the reliability of the reactor can be improved significantly. (Horiuchi, T.)

  1. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1981-01-01

    Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

  2. Control rod drive

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1986-01-01

    A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

  3. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  4. Seismic analysis of hydraulic control rod driving system

    International Nuclear Information System (INIS)

    Zheng, Yanhua; Bo, Hanliang; Dong, Duo

    2002-01-01

    A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)

  5. Fission reactor control rod

    International Nuclear Information System (INIS)

    Irie, Tomoo.

    1991-01-01

    The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)

  6. Regulatory perspective on incomplete control rod insertions

    International Nuclear Information System (INIS)

    Chatterton, M.

    1997-01-01

    The incomplete control rod insertions experienced at South Texas Unit 1 and Wolf Creek are of safety concern to the NRC staff because they represent potential precursors to loss of shutdown margin. Even before it was determined if these events were caused by the control rods or by the fuel there was an apparent correlation of the problem with high burnup fuel. It was determined that there was also a correlation between high burnup and high drag forces as well as with rod drop time histories and lack of rod recoil. The NRC staff initial actions were aimed at getting a perspective on the magnitude of the problem as far as the number of plants and the amount of fuel that could be involved, as well as the safety significance in terms of shutdown margin. As tests have been performed and data has been analyzed the focus has shifted more toward understanding the problem and the ways to eliminate it. At this time the staff's understanding of the phenomena is that it was a combination of factors including burnup, power history and temperature. The problem appears to be very sensitive to these factors, the interaction of which is not clearly understood. The model developed by Westinghouse provides a possible explanation but there is not sufficient data to establish confidence levels and sensitivity studies involving the key parameters have not been done. While several fixes to the problem have been discussed, no definitive fixes have been proposed. Without complete understanding of the phenomena, or fixes that clearly eliminate the problem the safety concern remains. The safety significance depends on the amount of shutdown margin lost due to incomplete insertion of the control rods. Were the control rods to stick high in the core, the reactor could not be shutdown by the control rods and other means such as emergency boration would be required

  7. Estimation of irradiated control rod worth

    International Nuclear Information System (INIS)

    Varvayanni, M.; Catsaros, N.; Antonopoulos-Domis, M.

    2009-01-01

    When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case that the level of detail that is known about the irradiation history of the control rod does not allow an accurate calculation of the absorber's burnup. The suggested methodology is based on measurements of the rod's worth carried out in the former core configuration and on corresponding calculations based on the original (before first irradiation) absorber concentration. The methodology is formulated for the general case of the multi-group theory; it is successfully tested for the one-group approximation, for a depleted control rod of the Greek Research Reactor, containing five neutron absorbers. The computations reproduce satisfactorily the irradiated rod worth measurements, practically eliminating the discrepancy of the total rod worth, compared to the computations based on the nominal absorber densities.

  8. Status of rod consolidation

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-04-01

    Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs

  9. Sucker rod string design of the pumping systems

    OpenAIRE

    Hua. L, C

    2015-01-01

    The existing design of sucker rod string mainly focuses on the simplifying assumptions that rod string was exposed to simple tension loading. And its goal was to have equal modified stress at the top of each taper. The improved rod design was to have the same degree of safety at each section, and it used a dynamic force distribution that was proportional along the whole string. Moreover, the available procedures did not provide the desired accuracy of its pertinent analysis, and the operators...

  10. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    International Nuclear Information System (INIS)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl

    2015-01-01

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs

  11. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  12. Development of absorber rod drive mechanisms for PFBR

    International Nuclear Information System (INIS)

    Veerasamy, R.; Dash, S.K.; Natarajan, S.; Rajan, M.; Prabhakar, R.; Kale, R.D.

    1997-01-01

    The Prototype Fast Breeder Reactor has two independent, diverse and fast acting shutdown systems each having its own neutron detectors, logic circuits, drive mechanisms and absorber rods. The respective drive mechanisms are called the control and safety rod drive mechanism and the diverse safety rod drive mechanism. The reliability of the shutdown systems has a direct bearing on the safety of the reactor. Hence a lot of development and testing efforts are required to optimise the design of the drive mechanisms and finally to qualify the same for reactor application. (author)

  13. Control rod shutdown system

    International Nuclear Information System (INIS)

    Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

    1996-01-01

    The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

  14. Why Rods and Cocci

    Indian Academy of Sciences (India)

    Bacteria exhibit a wide variety of shapes but the commonly studied species of bacteria are generally either spherical in shape which are called cocci (singular coccus) or have a cylindrical shape and are called rods or bacilli (singular bacillus). In reality rods and cocci are the ends of a continuum. Sonle of the cocci are.

  15. Control rod drives

    International Nuclear Information System (INIS)

    Oonuki, Koji.

    1981-01-01

    Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)

  16. Control-rod scram device

    International Nuclear Information System (INIS)

    Matsui, Yoshiro; Saito, Koji.

    1986-01-01

    Purpose: To eliminate the requirement for the nitrogen gas system in a scram device and enable safety and reliable shutdown of a water-cooled reactor power plant. Constitution: A piston and a spring are contained within a hydraulic vessel, and the piston is driven by the energy stored in the spring so as to supply hydraulic water to control mechanisms. During usual reactor operation, a scram valve is closed and a high water pressure of about 130 kg/cm 2 is applied to the water filled in the vessel through a check valve. Upon occurrence of abnormal conditions and generation of scram signals, the scram valve is opened to supply the water filled in the vessel through the scram valve to the control rod drive mechanisms. When the water pressure in the vessel is decreased, since the piston is urged upwardly by the energy stored in the spring, the water filled in the vessel is intermitently supplied to the control rod drive mechanisms. Thus, control rods can be inserted into the nuclear reactor to shutdown the same. (Horiuchi, T.)

  17. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  18. Control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Hiroyasu.

    1979-01-01

    Purpose: To enable rapid control in a simple circuit by providing a motor control device having an electric capacity capable of simultaneously driving all of the control rods rapidly only in the inserting direction as well as a motor controlling device capable of fine control for the insertion and extraction at usual operation. Constitution: The control rod drives comprise a first motor control device capable of finely controlling the control rods both in inserting and extracting directions, a second motor control device capable of rapidly driving the control rods only in the inserting direction, and a first motor switching circuit and a second motor switching circuit switched by switches. Upon issue of a rapid insertion instruction for the control rods, the second motor switching circuit is closed by the switch and the second motor control circuit and driving motors are connected. Thus, each of the control rod driving motors is driven at a high speed in the inserting direction to rapidly insert all of the control rods. (Yoshino, Y.)

  19. Multiple fuel rod gripper

    International Nuclear Information System (INIS)

    Shields, E.P.

    1987-01-01

    An apparatus is described for gripping an array of rods comprising: (a) gripping members grippingly engageable with the rods, each of which has a hollow portion terminating in an open end for receiving the end of one of the rods; (b) a closing means for causing the hollow portion of each of the gripping members to apply substantially the same gripping force onto the end of its respective rod, including (i) a locking plate having a plurality of tapered holes registrable with the array of rods, wherein the exterior of each of the gripping members is tapered and nested within one of the tapered holes, (ii) a withdrawing means having a hydraulic plunger operatively connected to each of the gripping members for applying a substantially identical withdrawing force on each of the gripping members, whereby the hollow portion of each of the gripping members applies substantially the same gripping force on its respective rod, and (c) means for detecting whether each of the gripping members has grippingly engaged its respective rod

  20. Fuel rod technology

    International Nuclear Information System (INIS)

    Bezold, H.; Romeiser, H.J.

    1979-07-01

    By extensive mechanization and automation of the fuel rod production, also at increasing production numbers, an efficient production shall be secured, simultaneously corresponding to the high quality standard of the fuel rods. The works done up to now concentrated on the lay out of a rough concept for a mechanized production course. Detail-studies were made for the problems of fuel rod humidity, filling and resistance welding. Further promotion of this project and thus further report will be stopped, since the main point of these works is the production technique. (orig.) [de

  1. Control rod testing apparatus

    International Nuclear Information System (INIS)

    Gaunt, R.R.; Ashman, C.M.

    1987-01-01

    A control rod testing apparatus is described comprising: a first guide means having a vertical cylindrical opening for grossly guiding a control rod; a second guide means having a vertical cylindrical opening for grossly guiding a control rod. The first and second guide means are supported at axially spaced locations with the openings coaxial; and a substantially cylindrical subassembly having a vertical cylindrical opening therethrough. The subassembly is trapped coaxial with and between the first and second guide means, and the subassembly radially floats with respect to the first and second guide means

  2. Analysis of control rod behavior based on numerical simulation

    International Nuclear Information System (INIS)

    Ha, D. G.; Park, J. K.; Park, N. G.; Suh, J. M.; Jeon, K. L.

    2010-01-01

    The main function of a control rod is to control core reactivity change during operation associated with changes in power, coolant temperature, and dissolved boron concentration by the insertion and withdrawal of control rods from the fuel assemblies. In a scram, the control rod assemblies are released from the CRDMs (Control Rod Drive Mechanisms) and, due to gravity, drop rapidly into the fuel assemblies. The control rod insertion time during a scram must be within the time limits established by the overall core safety analysis. To assure the control rod operational functions, the guide thimbles shall not obstruct the insertion and withdrawal of the control rods or cause any damage to the fuel assembly. When fuel assembly bow occurs, it can affect both the operating performance and the core safety. In this study, the drag forces of the control rod are estimated by a numerical simulation to evaluate the guide tube bow effect on control rod withdrawal. The contact condition effects are also considered. A full scale 3D model is developed for the evaluation, and ANSYS - commercial numerical analysis code - is used for this numerical simulation. (authors)

  3. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

    Purpose: To decrease the effect of water elimination and the effect of burn-up residue boron, thereby reduce the effect of burnable poison rods as the neutron poisons at the final stage of reactor core lifetime. Constitution: In a burnable poison rod according to the present invention, a hollow burnable poison material is filled in an external fuel can, an inner fuel can mounted with a carbon rod is inserted to the hollow portion of the burnable poison material and helium gases are charged in the outer fuel can. In such a burnable poison rod, the reactivity worths after the burning are reduced to one-half as compared with the conventional case. Accordingly, since the effect of the burnable poison as the neutron poisons is reduced at the final stage of the reactor core of lifetime, the excess reactivity of the reactor core is increased. (Horiuchi, T.)

  4. Thermal hydraulic performance of naturally aspirated control rod housing assemblies

    International Nuclear Information System (INIS)

    Geiger, G.T.; Randolph, H.W.; Paik, I.K.; Foti, D.J.

    1992-01-01

    Savannah River Site reactors are comprised of heat generating fuel/target assemblies, control rods which regulate reactor power, and heavy water which acts as the coolant and as a moderator. The fuel/target assemblies are cooled by the downflow of heavy water while the control rods are cooled via upflow. Five control rods are grouped with two safety rods in seven-channel assemblies called septifoils. Under normal operating conditions, the reactor power level, radial shape flux and axial power flux are regulated by the positioning of the control rods. The control rods are solid rods of a lithium-aluminum alloy with an thin aluminum outer sheath. Lithium is a good absorber of neutrons and, thus control rod temperatures rise with reactor power. At conditions of sufficiently high reactor power and degraded coolant flow, the control rods could heat sufficiently to cause a metallurigical failure of the sheath leading to molten material coming in contact with water and the possibility of a steam explosion. An accident has been postulated as part of the analysis involving the safety upgrade of Savannah River Site reactors in which the housing is not seated on the pin. Coolant from the upflow pin would not be directed into the housing but, into the moderator space surrounding the housing. Only naturally aspirated cooling due to buoyancy effects would be available to cool the control rods and the coolant mass flow rate would drop significantly from its nominal value. In this study, the mechanisms and limits of cooling heated rods housed in an unseated septifoil are addressed. Experiments were conducted on a shortened, prototypic housing with electrically heated rods to gain an understanding of the phenomena governing the cooling in such a case and develop data which can be used to evaluate predictive models. These experiments are described, their results discussed, and the predictions of current models is presented

  5. Control rod driving mechanism

    International Nuclear Information System (INIS)

    Ooshima, Yoshio.

    1983-01-01

    Purpose: To perform reliable scram operation, even if abnormality should occur in a system instructing scram operation in FBR type reactors. Constitution: An aluminum alloy member to be melt at a predetermined temperature (about 600sup(o)C) is disposed to a connection part between a control rod and a driving mechanism, whereby the control rod is detached from the driving mechanism and gravitationally fallen to the reactor core. (Ikeda, J.)

  6. Control rod drive

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1988-01-01

    Purpose: To provide a simple and economical control rod drive using a control circuit requiring no pulse circuit. Constitution: Control rods in a BWR type reactor are driven by hydraulic pressure and inserted or withdrawn in the direction of applying the hydraulic pressure. The direction of the hydraulic pressure is controlled by a direction control valve. Since the driving for the control rod is extremely important in view of the operation, a self diagnosis function is disposed for rapid inspection of possible abnormality. In the present invention, two driving contacts are disposed each by one between the both ends of a solenoid valve of the direction control valve for driving the control rod and the driving power source, and diagnosis is conducted by alternately operating them. Therefore, since it is only necessary that the control circuit issues a driving instruction only to one of the two driving contacts, the pulse circuit is no more required. Further, since the control rod driving is conducted upon alignment of the two driving instructions, the reliability of the control rod drive can be improved. (Horiuchi, T.)

  7. Method of inserting fuel rod

    International Nuclear Information System (INIS)

    Kamimoto, Shuji; Imoo, Makoto; Tsuchida, Kenji.

    1991-01-01

    The present invention concerns a method of inserting a fuel rod upon automatic assembling, automatic dismantling and reassembling of a fuel assembly in a light water moderated reactor, as well as a device and components used therefor. That is, a fuel rod is inserted reliably to an aimed point of insertion by surrounding the periphery of the fuel rod to be inserted with guide rods, and thereby suppressing the movement of the fuel rod during insertion. Alternatively, a fuel rod is inserted reliably to a point of insertion by inserting guide rods at the periphery of the point of insertion for the fuel rod to be inserted thereby surrounding the point of insertion with the guide rods or fuel rods. By utilizing fuel rods already present in the fuel assembly as the guide rods described above, the fuel rod can be inserted reliably to the point of insertion with no additional devices. Dummy fuel rods are previously inserted in a fuel assembly which are then utilized as the above-mentioned guide rods to accurately insert the fuel rod to the point of insertion. (I.S.)

  8. Safety quality classification test of the sealed neutron sources used in start-up neutron source rods for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Yao Chunbing; Guo Gang; Chao Jinglan; Duan Liming

    1992-01-01

    According to the regulations listed in the GB4075, the safety quality classification tests have been carried out for the neutron sources. The test items include temperature, external pressure, impact, vibration and puncture, Two dummy sealed sources are used for each test item. The testing equipment used have been examined and verified to be qualified by the measuring department which is admitted by the National standard Bureau. The leak rate of each tested sample is measured by UL-100 Helium Leak Detector (its minimum detectable leak rate is 1 x 10 -10 Pa·m 3 ·s -1 ). The samples with leak rate less than 1.33 x 10 -8 Pa·m 3 ·s -1 are considered up to the standard. The test results show the safety quality classification class of the neutron sources have reached the class of GB/E66545 which exceeds the preset class

  9. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  10. Control rod velocity limiter

    International Nuclear Information System (INIS)

    Cearley, J.E.; Carruth, J.C.; Dixon, R.C.; Spencer, S.S.; Zuloaga, J.A. Jr.

    1986-01-01

    This patent describes a velocity control arrangement for a reciprocable, vertically oriented control rod for use in a nuclear reactor in a fluid medium, the control rod including a drive hub secured to and extending from one end therefrom. The control device comprises: a toroidally shaped control member spaced from and coaxially positioned around the hub and secured thereto by a plurality of spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the toroidal member spaced therefrom in coaxial position. The side of the control member toward the control rod has a smooth generally conical surface. The side of the control member away from the control rod is formed with a concave surface constituting a single annular groove. The device also comprises inner and outer annular vanes radially spaced from one another and spaced from the side of the control member away from the control rod and positioned coaxially around and spaced from the hub and secured thereto by spaced radial webs thereby providing an annular passage for fluid intermediate the hub and the vanes. The vanes are angled toward the control member, the outer edge of the inner vane being closer to the control member and the inner edge of the outer vane being closer to the control member. When the control rod moves in the fluid in the direction toward the drive hub the vanes direct a flow of fluid turbulence which provides greater resistance to movement of the control rod in the direction toward the drive hub than in the other direction

  11. Method of driving control rod in reactor

    International Nuclear Information System (INIS)

    Osa, Hirotaka.

    1986-01-01

    Purpose: To improve security and safety of the reactor by reducing reactor output automatically and quickly when circulation of cooling water is stopped. Constitution: When the circulating pump is under operation, fluid pressure in the discharge pipe is transferred to the fluid room of fluid pressure cylinder via the control rod drive pipe and lift up the piston, and then the control rod is drawn out of the reactor core. When the circulating pump is lowered in its functions, discharge pipe fluid pressure decreases, fluid pressure in the fluid room decreases, and with less force of piston movement, the control rod gets lowered by its own weight. At this time, the blocked state of the opening by the piston is released, fluid flows into the room. Lowering of pressure and the control rod is promoted by transferring out fluid below the piston in the fluid room to the upper part of the piston via a small gap when the control rod falls by gravity. (Horiuchi, T.)

  12. Control rod housing alignment

    International Nuclear Information System (INIS)

    Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

    1990-01-01

    This patent describes a process for measuring the vertical alignment between a hole in a core plate and the top of a corresponding control rod drive housing within a boiling water reactor. It comprises: providing an alignment apparatus. The alignment apparatus including a lower end for fitting to the top of the control rod drive housing; an upper end for fitting to the aperture in the core plate, and a leveling means attached to the alignment apparatus to read out the difference in angularity with respect to gravity, and alignment pin registering means for registering to the alignment pin on the core plate; lowering the alignment device on a depending support through a lattice position in the top guide through the hole in the core plate down into registered contact with the top of the control rod drive housing; registering the upper end to the sides of the hole in the core plate; registering the alignment pin registering means to an alignment pin on the core plate to impart to the alignment device the required angularity; and reading out the angle of the control rod drive housing with respect to the hole in the core plate through the leveling devices whereby the angularity of the top of the control rod drive housing with respect to the hole in the core plate can be determined

  13. SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  14. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  15. Hydraulically centered control rod

    International Nuclear Information System (INIS)

    Horlacher, W.R.; Sampson, W.T.; Schukei, G.E.

    1981-01-01

    A control rod suspended to reciprocate in a guide tube of a nuclear fuel assembly has a hydraulic bearing formed at its lower tip. The bearing includes a plurality of discrete pockets on its outer surface into which a flow of liquid is continuously provided. In one embodiment the flow is induced by the pressure head in a downward facing chamber at the end of the bearing. In another embodiment the flow originates outside the guide tube. In both embodiments the flow into the pockets produces pressure differences across the bearing which counteract forces tending to drive the rod against the guide tube wall. Thus contact of the rod against the guide tube is avoided

  16. Control rod control device

    International Nuclear Information System (INIS)

    Seiji, Takehiko; Obara, Kohei; Yanagihashi, Kazumi

    1998-01-01

    The present invention provides a device suitable for switching of electric motors for driving each of control rods in a nuclear reactor. Namely, in a control rod controlling device, a plurality of previously allotted electric motors connected in parallel as groups, and electric motors of any selected group are driven. In this case, a voltage of not driving predetermined selected electric motors is at first applied. In this state an electric current supplied to the circuit of predetermined electric motors is detected. Whether integration or failure of a power source and the circuit of the predetermined electric motors are normal or not is judged by the detected electric current supplied. After they are judged normal, the electric motors are driven by a regular voltage. With such procedures, whether the selected circuit is normal or not can be accurately confirmed previously. Since the electric motors are not driven just at the selected time, the control rods are not operated erroneously. (I.S.)

  17. Burnable poison rod

    International Nuclear Information System (INIS)

    Natsume, Tomohiro.

    1988-01-01

    Purpose: To increase the reactor core lifetime by decreasing the effect of neutron absorption of burnable poison rods by using material with less neutron absorbing effect. Constitution: Stainless steels used so far as the coating material for burnable poison rods have relatively great absorption in the thermal neutral region and are not preferred in view of the neutron economy. Burnable poison rods having fuel can made of zirconium alloy shows absorption the thermal neutron region lower by one digit than that of stainless steels but they shows absorption in the resonance region and the cost is higher. In view of the above, the fuel can of the burnable poison material is made of aluminum or aluminu alloy. This can reduce the neutron absorbing effect by stainless steel fuel can and effectively utilize neutrons that have been wastefully absorbed and consumed in stainless steels. (Takahashi, M.)

  18. BWR control rod drive scram pilot valve monitoring system

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1984-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechancial works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the ''insert'' side of the control rod piston and vents the ''withdraw'' side of the piston causing the rods to insert during a scam. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a ''half scram'', a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  19. BWR control rod drive scram pilot valve monitoring program

    International Nuclear Information System (INIS)

    Soden, R.A.; Kelly, V.

    1986-01-01

    The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechanical works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the insert side of the control rod piston and vents the withdraw side of the piston causing the rods to insert during a scram. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a half scram, a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

  20. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

    For power upgrading VVER-440 reactors we need to know exactly how the temperature measured by the thermocouples is related to the average outlet temperature of the fuel assemblies. Accordingly, detailed knowledge on mixing process in the rod bundles and in the fuel assembly head have great importance. Here we study the hydrodynamics of rod bundles based on the results of direct numerical and large eddy simulation of flows in subchannels. It is shown that secondary flow and flow pulsation phenomena can be observed using both methodologies. Some consequences of these observations are briefly discussed. (author)

  1. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Hernborg, G; Flinta, J E

    1964-08-15

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within {+-} 5 per cent.

  2. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Hernborg, G.; Flinta, J.E.

    1964-08-01

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within ± 5 per cent

  3. Proposal of a synchro panel meter instrument to replace the obsolete Synchro/Resolver reading device used as position indicator of safety rods assembly of the Brazilian IEA-R1 Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Toledo, Fabio de; Brancaccio, Franco; Cardenas, Jose Patricio N.

    2015-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) was founded in 1956 (as Atomic Energy Institute - IEA) as a facility complex, for the research, development and application, in the nuclear technology field. The institute is recognized as a national leader in nuclear research and development (R and D), including the areas of reactor operation, radiopharmaceuticals, industrial and laboratory applications, materials science and laser technologies and applications. IPEN's main facility is the IEA-R1, nuclear research reactor (NRR), today, the only one in Brazil with a power level suitable for applications in physics, chemistry, biology and engineering. Some radioisotopes are also produced in IEA-R1, for medical and other applications. A common problem faced in the IEA-R1 maintenance is instrumentation obsolescence; spare parts are no more available, because of discontinued production, and an updating program is mandatory, aiming at modernization of old-aged I and C systems. In the presented context, an electronic system is here proposed, as a replacement for the reactor safety (shim) rods assembly position indicator, based on an open-source physical computing platform called Arduino, which includes a simple microcontroller board and a software-code development environment. A mathematical algorithm for the synchro-motor signal processing was developed, and the obtained resolution was better than 1.5%. (author)

  4. Proposal of a synchro panel meter instrument to replace the obsolete Synchro/Resolver reading device used as position indicator of safety rods assembly of the Brazilian IEA-R1 Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Toledo, Fabio de; Brancaccio, Franco; Cardenas, Jose Patricio N., E-mail: fatoledo@ipen.br, E-mail: fbrancac@ipen.br, E-mail: ahiru@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) was founded in 1956 (as Atomic Energy Institute - IEA) as a facility complex, for the research, development and application, in the nuclear technology field. The institute is recognized as a national leader in nuclear research and development (R and D), including the areas of reactor operation, radiopharmaceuticals, industrial and laboratory applications, materials science and laser technologies and applications. IPEN's main facility is the IEA-R1, nuclear research reactor (NRR), today, the only one in Brazil with a power level suitable for applications in physics, chemistry, biology and engineering. Some radioisotopes are also produced in IEA-R1, for medical and other applications. A common problem faced in the IEA-R1 maintenance is instrumentation obsolescence; spare parts are no more available, because of discontinued production, and an updating program is mandatory, aiming at modernization of old-aged I and C systems. In the presented context, an electronic system is here proposed, as a replacement for the reactor safety (shim) rods assembly position indicator, based on an open-source physical computing platform called Arduino, which includes a simple microcontroller board and a software-code development environment. A mathematical algorithm for the synchro-motor signal processing was developed, and the obtained resolution was better than 1.5%. (author)

  5. Device for coupling a control rod and control rod drive

    International Nuclear Information System (INIS)

    Nishioka, Kazuya.

    1975-01-01

    Object: To obtain simple and reliable coupling between a control rod and control rod drive by equipping the lower end of the control rod with an extension provided with lateral protuberances and forming the upper end of an index tube with a recess provided with lateral holes. Structure: The tapering central extension of the control rod is inserted into the recess by lowering the control rod, and then it is further inserted by causing frictional movement of the inclined surfaces of lateral protuberances in frictional contact with guide surfaces. When the lateral protuberances are brought into contact with a stepped portion, the control rod is rotated to fit the lateral protuberances into the lateral holes. In this way, the control rod is coupled to the index tube of the control rod drive. (Yoshino, Y.)

  6. Study on shadowing effect caused by transient rods at NSRR

    International Nuclear Information System (INIS)

    Nakamura, T.; Yachi, S.; Ishijima, K.

    1992-01-01

    Irregularly inserted three control rods created so called shadowing effects on some of the neutronic instruments at the Nuclear Safety Research Reactor (NSRR). During operations at the reactor power of up to 10 MW, the three control rods called transient rods, could be fully or partly inserted into the NSRR core. Reactor power monitors located outside of the core at the direction of deeply inserted transient rods indicated lower power in such operations. Power profiles of the reactor and neutron fluxes at power monitor locations were calculated with a three dimensional neutron diffusion code, CITATION. The calculation indicated that the real reactor power could be smaller than the measured maximum power by as mush as 30 % in such operations. The calculated neutron fluxes well described the changes in the apparent power monitor indications as a function of the transient rod position. (author)

  7. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  8. Control rod drives

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1986-01-01

    Purpose: To enable to direct disconnection of control rods upon abnormal temperature rise in the reactor thereby improve the reliability for the disconnecting operation in control rod drives for FBR type reactors upon emergency. Constitution: A diaphragm is disposed to the upper opening of a sealing vessel inserted to the hollow portion of an electromagnet and a rod is secured to the central position of the upper surface. A spring contacts are attached by way of an insulator to the inner surface at the lower portion of an extension pipe and connected with cables for supplying electric power sources respectively to a magnet. If the temperature in the reactor abnormally rises, liquid metals in the sealing vessel are expanded tending to extend the bellows downwardly. However, since they are attracted by the electromagnet, the thermal expansion of the liquid metals exert on the diaphragm prior to the bellows. Thus, the switch between the spring contacts is made open to attain the deenergized state to thereby disconnect the control rod and shutdown the neclear reactor. (Horiuchi, T.)

  9. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Mizuno, Katsuyuki.

    1976-01-01

    Object: To restrict the reduction in performance due to stress corrosion cracks by making use of condensate produced in a turbine steam condenser. Structure: Water produced in a turbine steam condenser is forced into a condensed water desalting unit by low pressure condensate pump. The condensate is purified and then forced by a high pressure condensate pump into a feedwater heater for heating before it is returned to the reactor by a feedwater pump. Part of the condensate issuing from the condensate desalting unit is branched from the remaining portion at a point upstream the pump and is withdrawn into a control rod drive water pump after passing through a motordriven bypass valve, an orifice and a condenser water level control valve, is pressurized in the control rod drive water desalting unit and supplied to a control rod drive water pressure system. The control rod is vertically moved by the valve operation of the water pressure system. Since water of high oxygen concentration does not enter during normal operation, it is possible to prevent the stress cracking of the stainless steel apparatus. (Nakamura, S.)

  10. Trunnion Rod Microcrack Detection

    Science.gov (United States)

    2013-08-01

    Richard W. Haskins, Joseph A. Padula , and John E. Hite BACKGROUND: Post-tensioned rods are used to anchor spillway gates and transfer the forces...email: James.A.Evans@usace.army.mil). This technical note should be cited as follows: Evans, J. A., Haskins, R. W., Padula , J. A., and Hite, J. E. 2013

  11. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  12. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

    Moulton, D.E.

    2013-02-01

    A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

  13. Control rod cluster with removable rods for nuclear fuel assembly

    International Nuclear Information System (INIS)

    Denizou, J.P.

    1989-01-01

    For each removable control rod, the open end section of the sleeve has a certain length of reduced diameter with openings in its wall. The top end of the rod is joined to an extension tube that surrounds the shaft over part of its lenght. This extension tube fits over the reduced part of the sleeve when the shaft is screwed into the bore of the sleeve. Rotation of the rod in the sleeve is prevented by deforming the extension tube locally in the openings of the end part of the sleeve. The rod is dismantled by exerting a torque on it using a gripping area near the end of the rod [fr

  14. Morphoelastic rods. Part I: A single growing elastic rod

    KAUST Repository

    Moulton, D.E.; Lessinnes, T.; Goriely, A.

    2013-01-01

    A theory for the dynamics and statics of growing elastic rods is presented. First, a single growing rod is considered and the formalism of three-dimensional multiplicative decomposition of morphoelasticity is used to describe the bulk growth of Kirchhoff elastic rods. Possible constitutive laws for growth are discussed and analysed. Second, a rod constrained or glued to a rigid substrate is considered, with the mismatch between the attachment site and the growing rod inducing stress. This stress can eventually lead to instability, bifurcation, and buckling. © 2012 Elsevier Ltd. All rights reserved.

  15. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  16. Control rod drive shaft latch

    International Nuclear Information System (INIS)

    Thorp, A.G. II.

    1976-01-01

    A latch mechanism is operated by differential pressure on a piston to engage the drive shaft for a control rod in a nuclear reactor, thereby preventing the control rod from being ejected from the reactor in case of failure of the control rod drive mechanism housing which is subjected to the internal pressure in the reactor vessel. 6 claims, 4 drawing figures

  17. Hydraulic system for the drive of control rod

    International Nuclear Information System (INIS)

    Niwano, Masao.

    1978-01-01

    Purpose: To remove thermal stress and improve safety by utilizing water discharged a driving device as a part of cooling water for the device upon driving of control rods. Constitution: A water drain valve is wholly closed and a flow stabilization valve is supplied with an amount of water necessary for driving control rods. Upon driving one control rod, an amount of water required for the driving is caused to flow to the relivant hydraulic control unit and the flow rate in the stabilization valve is reduced by an amount required for the driving to keep the flow rate constant in the flow control valve. Since Excess water conventionally returned to the pressure vessel is utilized as cooling water for the driving device of control rods, the pressure vessel nozzle can be saved. Accordingly, the thermal stress in the nozzle portion can be removed to significantly improve the safety. (Seki, T.)

  18. Physics analysis of the gang partial rod drive event

    International Nuclear Information System (INIS)

    Boman, C.; Frost, R.L.

    1992-08-01

    During the routine positioning of partial-length control rods in Gang 3 on the afternoon of Monday, July 27, 1992, the partial-length rods continued to drive into the reactor even after the operator released the controlling toggle switch. In response to this occurrence, the Safety Analysis and Engineering Services Group (SAEG) requested that the Applied Physics Group (APG) analyze the gang partial rod drive event. Although similar accident scenarios were considered in analysis for Chapter 15 of the Safety Analysis Report (SAR), APG and SAEG conferred and agreed that this particular type of gang partial-length rod motion event was not included in the SAR. This report details this analysis

  19. Investigation of axial power gradients near a control rod tip

    Energy Technology Data Exchange (ETDEWEB)

    Loberg, John, E-mail: John.Loberg@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Osterlund, Michael, E-mail: Michael.Osterlund@fysast.uu.se [Uppsala University, Department of Physics and Astronomy, Division of Applied Nuclear Physics, Box 525, SE-75120 Uppsala (Sweden); Bejmer, Klaes-Hakan, E-mail: Klaes-Hakan.Bejmer@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Blomgren, Jan, E-mail: Jan.Blomgren@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden); Kierkegaard, Jesper, E-mail: Jesper.Kierkegaar@vattenfall.com [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, 162 60 Vaellingby, Stockholm (Sweden)

    2011-07-15

    Highlights: > Pin power gradients near BWR control rod tips have been investigated. > A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. > Small nodes increases pin power gradients; standard nodes underestimates gradients. > The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, {approx}15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  20. Investigation of axial power gradients near a control rod tip

    International Nuclear Information System (INIS)

    Loberg, John; Osterlund, Michael; Bejmer, Klaes-Hakan; Blomgren, Jan; Kierkegaard, Jesper

    2011-01-01

    Highlights: → Pin power gradients near BWR control rod tips have been investigated. → A control rod tip is modeled in MCNP and compared to simplified 2D/3D geometry. → Small nodes increases pin power gradients; standard nodes underestimates gradients. → The MCNP results are validated against axial gamma scan of a controlled fuel pin. - Abstract: Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, ∼15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  1. Control rod calibration including the rod coupling effect

    International Nuclear Information System (INIS)

    Szilard, R.; Nelson, G.W.

    1984-01-01

    In a reactor containing more than one control rod, which includes all reactors licensed in the United States, there will be a 'coupling' or 'shadowing' of control rod flux at the location of a control rod as a result of the flux depression caused by another control rod. It was decided to investigate this phenomenon further, and eventually to put calibration table data or formulae in a small computer in the control room, so once could insert the positions of the three control rods and receive the excess reactivity without referring to separate tables. For this to be accomplished, a 'three control- rod reactivity function' would be used which would include the flux coupling between the rods. The function is design and measured data was fitted into it to determine the calibration constants. The input data for fitting the trial functions consisted of 254 data points, each consisting of the position of the reg, shim, and transient rods, and the total excess reactivity. (About 200 of these points were 'critical balance points', that is the rod positions for which reactor was critical, and the remainder were determined by positive period measurements.) Although this may be unrealistic from a physical viewpoint, the function derived gave a very accurate recalculation of the input data, and thus would faithfully give the excess reactivity for any possible combination of the locations of the three control rods. The next step, incorporation of the three-rod function into the minicomputer, will be pursued in the summer and fall of 1984

  2. Control rod studies in small and medium sized fast reactors

    International Nuclear Information System (INIS)

    John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

    1988-01-01

    Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

  3. COMETHE III-M for transient fuel rod behaviour prediction

    International Nuclear Information System (INIS)

    Billaux, M.; Vliet, J. van

    1983-01-01

    The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

  4. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

  5. Computer program for automatic generation of BWR control rod patterns

    International Nuclear Information System (INIS)

    Taner, M.S.; Levine, S.H.; Hsia, M.Y.

    1990-01-01

    A computer program named OCTOPUS has been developed to automatically determine a control rod pattern that approximates some desired target power distribution as closely as possible without violating any thermal safety or reactor criticality constraints. The program OCTOPUS performs a semi-optimization task based on the method of approximation programming (MAP) to develop control rod patterns. The SIMULATE-E code is used to determine the nucleonic characteristics of the reactor core state

  6. Fuel rod fixing system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    This is a reusable system for fixing a nuclear reactor fuel rod to a support. An interlock cap is fixed to the fuel rod and an interlock strip is fixed to the support. The interlock cap has two opposed fingers, which are shaped so that a base is formed with a body part. The interlock strip has an extension, which is shaped so that this is rigidly fixed to the body part of the base. The fingers of the interlock cap are elastic in bending. To fix it, the interlock cap is pushed longitudinally on to the interlock strip, which causes the extension to bend the fingers open in order to engage with the body part of the base. To remove it, the procedure is reversed. (orig.) [de

  7. Fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member or vice versa. The locking cap has two opposing fingers and shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed. In an alternative embodiment, the cap is rigid and the strip is transversely resiliently compressible. (author)

  8. Control rod drives

    International Nuclear Information System (INIS)

    Yamanaka, Toshikatsu.

    1979-01-01

    Purpose: To protect bellows against failures due to negative pressure to prevent the loss of pressure balance caused by the expansion of the bellows upon scram. Constitution: An expansion pipe connected to the control rod drive is driven along a guide pipe to insert a control rod into the reactor core. Expansible bellows are provided at the step between the expansion pipe and the guide pipe. Further, a plurality of bore holes or slits are formed on the side wall of the guide pipe corresponding to the expansion portion of the bellows. In such an arrangement, when the expansion pipe falls rapidly and the bellows are expanded upon scram, the volume between each of the pipes of the bellows and the guide pipe is increased to produce a negative pressure, but the effect of the negative pressure on the bellows can be eliminated by the flowing-in of coolants corresponding to that pressure through the bore holes or the slits. (Furukawa, Y.)

  9. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  10. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  11. Control rod withdrawal monitoring device

    International Nuclear Information System (INIS)

    Ebisuya, Mitsuo.

    1984-01-01

    Purpose: To prevent the power ramp even if a plurality of control rods are subjected to withdrawal operation at a time, by reducing the reactivity applied to the reactor. Constitution: The control rod withdrawal monitoring device is adapted to monitor and control the withdrawal of the control rods depending on the reactor power and the monitoring region thereof is divided into a control rod group monitoring region a transition region and a control group monitoring not interfere region. In a case if the distance between a plurality of control rods for which the withdrawal positions are selected is less than a limiting value, the coordinate for the control rods, distance between the control rods and that the control rod distance is shorter are displayed on a display panel, and the withdrawal for the control rods are blocked. Accordingly, even if a plurality of control rods are subjected successively to the withdrawal operation contrary to the control rod withdrawal sequence upon high power operation of the reactor, the power ramp can be prevented. (Kawakami, Y.)

  12. Rod drive and latching mechanism

    International Nuclear Information System (INIS)

    Veronesi, L.; Sherwood, D.G.

    1982-01-01

    Hydraulic drive and latching mechanisms for driving reactivity control mechanisms in nuclear reactors are described. Preferably, the pressurized reactor coolant is utilized to raise the drive rod into contact with and to pivot the latching mechanism so as to allow the drive rod to pass the latching mechanism. The pressure in the housing may then be equalized which allows the drive rod to move downwardly into contact with the latching mechanism but to hold the shaft in a raised position with respect to the reactor core. Once again, the reactor coolant pressure may be utilized to raise the drive rod and thus pivot the latching mechanism so that the drive rod passes above the latching mechanism. Again, the mechanism pressure can be equalized which allows the drive rod to fall and pass by the latching mechanism so that the drive rod approaches the reactor core. (author)

  13. The experimental development and performance test of the pneumatic control-rod drive for the THTR

    International Nuclear Information System (INIS)

    Lange, G.; Boehlo, D.; Heim, H.; Kleine-Tebbe, A.

    1976-01-01

    Reactor control and shutdown of the THTR is accomplished by two independent systems, the first consisting of 36 absorber rods penetrating the graphite reflector region surrounding the core, the second consisting of 42 absorber rods that insert directly into the pebble bed core. This paper describes the design development and testing of the pneumatic rod drives used for movement of the 42 core control rods. The core control rods have two functions: the first, for reactor safety purposes, provides for adequate safe shutdown of the reactor under cold conditions; the second, for operational purposes, provides for compensation of slow changes in reactivity. The safety and operational functions for each absorber rod are respectively carried out by a long-stroke-piston pneumatic drive and by a stepping-piston pneumatic drive, both of these independent, helium-driven drives being incorporated in the rod drive unit for each control rod. To study the performance of the rod drive, a complete prototype control rod and rod drive unit was built and tested under simulated reactor operational conditions. Operational experience under helium temperatures and pressures was gained and the drives were tested under stress and simulated accident conditions. The reliability of this system has been demonstrated to licensing authorities and to the customer. The programme will be completed with the commissioning tests of drives for the THTR-300 reactor. (author)

  14. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  15. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    To ensure fuel design safety and structural integrity requires the response prediction of fuel rod to reactor coolant flow excitation. However, there are many obstacles in predicting the response as described. Even if the response can be predicted, the design criteria on wear failure, including correlation with the vibration, may be difficult to establish because of a variety of related parameters, such as material, surface condition and environmental factors. Thus, a prototype test for each new fuel assembly design, i.e. a long-term endurance test, is performed for design validation with respect to flow-induced vibration (FIV) and wear. There are still needs of theoretical prediction methods for the response and anticipated failure. This paper revisits the general aspect on the response prediction, mathematical description, analysis procedure and wear correlation aspect of fuel rod's FIV

  16. Cone rod dystrophies

    Science.gov (United States)

    Hamel, Christian P

    2007-01-01

    Cone rod dystrophies (CRDs) (prevalence 1/40,000) are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP), also called the rod cone dystrophies (RCDs) resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7). Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far). The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs), CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs), and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs). It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is always advised. Currently

  17. Cone rod dystrophies

    Directory of Open Access Journals (Sweden)

    Hamel Christian P

    2007-02-01

    Full Text Available Abstract Cone rod dystrophies (CRDs (prevalence 1/40,000 are inherited retinal dystrophies that belong to the group of pigmentary retinopathies. CRDs are characterized by retinal pigment deposits visible on fundus examination, predominantly localized to the macular region. In contrast to typical retinitis pigmentosa (RP, also called the rod cone dystrophies (RCDs resulting from the primary loss in rod photoreceptors and later followed by the secondary loss in cone photoreceptors, CRDs reflect the opposite sequence of events. CRD is characterized by primary cone involvement, or, sometimes, by concomitant loss of both cones and rods that explains the predominant symptoms of CRDs: decreased visual acuity, color vision defects, photoaversion and decreased sensitivity in the central visual field, later followed by progressive loss in peripheral vision and night blindness. The clinical course of CRDs is generally more severe and rapid than that of RCDs, leading to earlier legal blindness and disability. At end stage, however, CRDs do not differ from RCDs. CRDs are most frequently non syndromic, but they may also be part of several syndromes, such as Bardet Biedl syndrome and Spinocerebellar Ataxia Type 7 (SCA7. Non syndromic CRDs are genetically heterogeneous (ten cloned genes and three loci have been identified so far. The four major causative genes involved in the pathogenesis of CRDs are ABCA4 (which causes Stargardt disease and also 30 to 60% of autosomal recessive CRDs, CRX and GUCY2D (which are responsible for many reported cases of autosomal dominant CRDs, and RPGR (which causes about 2/3 of X-linked RP and also an undetermined percentage of X-linked CRDs. It is likely that highly deleterious mutations in genes that otherwise cause RP or macular dystrophy may also lead to CRDs. The diagnosis of CRDs is based on clinical history, fundus examination and electroretinogram. Molecular diagnosis can be made for some genes, genetic counseling is

  18. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  19. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  20. Apparatus for inspecting a irradiated nuclear fuel rod

    International Nuclear Information System (INIS)

    Saura, Hideaki; Yonemura, Eizo.

    1975-01-01

    Object: To increase safety and inspection efficiency by operating irradiated fuel rods, which are accommodated in a water-filled pool after being taken out from the reactor. Structure: When making inspection of irradiated fuel rods, particularly the cladding tube thereof, a fuel box which stores irradiated fuel rods in a water pool is secured to a securement mechanism with slime removal apparatus and inspection apparatus on either side capable of being vertically moved, and it is then stopped at a water depth of about 2 meters. When the lid of the box is opened, irradiated fuel rods are taken out with gripping means and then secured together with the gripping means to an operation base provided on the outside of the pool. Thereafter, the box is lowered by operating pedals on the operation base to completely pull out the irradiated fuel rods from the box, and the irradiated fuel rods are then horizontally moved and then held in a suspended state. Next a slime removal apparatus in raised by operating pedals and an inspection element assembly are progressively raised for inspection of the state of the cladding tube of each fuel rod after removal of slime therefrom. (Nakamura, S.)

  1. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  2. Theoretical calibration of grey and black control rods of gas-graphite power reactors

    International Nuclear Information System (INIS)

    Joksimovic, V.

    1964-01-01

    Full text: Calculation of calibration curve for particular control rod batches is of significant importance for safety and operation reasons. The procedure presented in this paper is based on the two following criteria: Constants of the lattice region with control rods are determined by supercell method. Effective multiplication constant of the core dependent on the insertion of control rods was determined by dividing the core onto two axial and radial zones. Calculation of the black control rods takes into account epithermal absorption. Thermal extrapolated length of the control rods was calculated by using Kushneriuk-McKey relation. The extrapolated length of the grey rods and the epithermal extrapolated length of the black rods were calculated by diffusion theory. Correlation procedure was used for calculation of epithermal extrapolated length. The complete mathematical procedure was programmed for calculations on the digital ZUSE-Z-23 and ELLIOTT-803-B computers

  3. Force analysis of the advanced neutron source control rod drive latch mechanism

    International Nuclear Information System (INIS)

    Damiano, B.

    1989-01-01

    The Advanced Neutron Source reactor (ANS), a proposed Department of Energy research reactor currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL), will generate a thermal neutron flux approximating 10 30 M -2 emdash S -1 . The compact core necessary to produce this flux provides little space for the shim safety control rods, which are located in the central annulus of the core. Without proper control rod drive design, the control rod drive magnets (which hold the control rod latch in a ready-to-scram position) may be unable to support the required load due to their restricted size. This paper describes the force analysis performed on the control rod latch mechanism to determine the fraction of control rod weight transferred to the drive magnet. This information will be useful during latch, control rod drive and magnet design. 5 refs., 12 figs

  4. Fuel rod pellet loading head

    International Nuclear Information System (INIS)

    Howell, T.E.

    1975-01-01

    An assembly for loading nuclear fuel pellets into a fuel rod comprising a loading head for feeding pellets into the open end of the rod is described. The pellets rest in a perforated substantially V-shaped seat through which air may be drawn for removal of chips and dust. The rod is held in place in an adjustable notched locator which permits alignment with the pellets

  5. Control-rod driving mechanism

    International Nuclear Information System (INIS)

    Jodoi, Takashi.

    1976-01-01

    Purpose: To prevent falling of control rods due to malfunction. Constitution: The device of the present invention has a scram function in particular, and uses principally a fluid pressure as a scram accelerating means. The control rod is held by upper and lower holding devices, which are connected by a connecting mechanism. This connecting mechanism is designed to be detachable only at the lower limit of driving stroke of the control rod so that there occurs no erroneous scram resulting from careless disconnection of the connecting mechanism. Further, scramming operation due to own weight of the scram operating portion such as control rod driving shaft may be effected to increase freedom. (Kamimura, M.)

  6. Integrated control rod monitoring device

    International Nuclear Information System (INIS)

    Saito, Katsuhiro

    1997-01-01

    The present invention provides a device in which an entire control rod driving time measuring device and a control rod position support device in a reactor building and a central control chamber are integrated systematically to save hardwares such as a signal input/output device and signal cables between boards. Namely, (1) functions of the entire control rod driving time measuring device for monitoring control rods which control the reactor power and a control rod position indication device are integrated into one identical system. Then, the entire devices can be made compact by the integration of the functions. (2) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated in a central operation board and a board in the site. Then, the place for the installation of them can be used in common in any of the cases. (3) The functions of the entire control rod driving time measuring device and the control rod position indication device are integrated to one identical system to save hardware to be used. Then, signal input/output devices and drift branching panel boards in the site and the central operation board can be saved, and cables for connecting both of the boards is no more necessary. (I.S.)

  7. Reconstitutable control rod spider assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferian, S.J.

    1990-01-01

    A reconstitutable control rod/spider assembly includes a hollow connecting finger of the spider having a pair of opposing flat segments formed on the interior thereof and engaging a pair of opposing flat sectors formed on the exterior of a stem extending form the upper end of control rod. The stem also has an externally-threaded portion engaging a nut and a pilot aligning portion for the nut. The nut has a radially flexible and expandable thread-defining element captured in its bore. The segments and sectors allow the rod to be removed and reattached after turning through 180 0 to allow more even wear on the rod. (author)

  8. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

  9. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

  10. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  11. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Wada, Toyoji.

    1979-01-01

    Purpose: To remove failures caused from combination of fuel-cladding interactions, hydrogen absorptions, stress corrosions or the likes by setting the quantity ratio of uranium or uranium and plutonium relative to oxygen to a specific range in fuel pellets and forming a specific size of a through hole at the center of the pellets. Constitution: In a fuel rods of a structure wherein fuel pellets prepared by compacting and sintering uranium dioxide, or oxide mixture consisting of oxides of plutonium and uranium are sealed with a zirconium metal can, the ratio of uranium or uranium and plutonium to oxygen is specified as 1 : 2.01 - 1 : 2.05 in the can and a passing hole of a size in the range of 15 - 30% of the outer diameter of the fuel pellet is formed at the center of the pellet. This increases the oxygen partial pressure in the fuel rod, oxidizes and forms a protection layer on the inner surface of the can to control the hydrogen absorption and stress corrosion. Locallized stress due to fuel cladding interaction (PCMI) can also be moderated. (Horiuchi, T.)

  12. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  13. Thermalhydraulic phenomena governing the quenching of hot rods, and existing models

    International Nuclear Information System (INIS)

    Bestion, D.

    2001-01-01

    After a core dry-out and a period of rod clad overheating, which might occur in some postulated accidental sequences in a PWR, the actuation of safety injections allows to quench the hot rods. Both thermal and mechanical processes control the phenomenon of quenching. Quenching first requires that liquid water is present to release the heat stored in the rod. When water is present, a pre-cooling of the clad is also required before quenching. (author)

  14. Thermalhydraulic phenomena governing the quenching of hot rods, and existing models

    Energy Technology Data Exchange (ETDEWEB)

    Bestion, D. [CEA-Grenoble, DEN/DTP/SMTH (France)

    2001-07-01

    After a core dry-out and a period of rod clad overheating, which might occur in some postulated accidental sequences in a PWR, the actuation of safety injections allows to quench the hot rods. Both thermal and mechanical processes control the phenomenon of quenching. Quenching first requires that liquid water is present to release the heat stored in the rod. When water is present, a pre-cooling of the clad is also required before quenching. (author)

  15. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    International Nuclear Information System (INIS)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor

  16. Critical heat flux detection in rods simulating fuel elements by using dilation method

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1993-01-01

    In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

  17. Nuclear safety - Topical issues

    International Nuclear Information System (INIS)

    1995-01-01

    The following topical issues related to nuclear safety are discussed: steam generators; maintenance strategies; control rod drive nozzle cracks; core shrouds cracks; sump strainer blockage; fire protection; computer software important for safety; safety during shutdown; operational safety experience; external hazards and other site related issues. 5 figs, 5 tabs

  18. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

    The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

  19. Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

    International Nuclear Information System (INIS)

    Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

    1999-06-01

    Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

  20. Digital control rod blocking monitor

    International Nuclear Information System (INIS)

    Funayama, Yoshio.

    1996-01-01

    The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

  1. Cuisenaire Rods Go to College.

    Science.gov (United States)

    Chinn, Phyllis; And Others

    1992-01-01

    Presents examples of questions and answers arising from a hands-on and exploratory approach to discrete mathematics using cuisenaire rods. Combinatorial questions about trains formed of cuisenaire rods provide the setting for discovering numerical patterns by experimentation and organizing the results using induction and successive differences.…

  2. Control rod experiments in Racine

    International Nuclear Information System (INIS)

    Stanculescu, A.; Humbert, G.

    1981-09-01

    A survey of the control-rod experiments planned within the joint CEA/CNEN-DeBeNe critical experiment RACINE is given. The applicability to both heterogeneous and homogeneous large power LMFBR-cores is discussed. Finally, the most significant results of the provisional design calculations performed on behalf of the RACINE control-rod programme are presented

  3. Control rod drives

    International Nuclear Information System (INIS)

    Furumitsu, Yutaka.

    1981-01-01

    Purpose: To improve the reliability of a device for driving an LMFBR type reactor control rod by providing a buffer unit having a stationary electromagnetic coil and a movable electromagnetic coil in the device to thereby avord impact stress at scram time and to simplify the structure of the buffer unit. Constitution: A non-contact type buffer unit is constructed with a stationary electromagnetic coil, a cable for the stationary coil, a movable electromagnetic coil, a spring cable for the movable coil, and a backup coil spring or the like. Force produced at scram time is delivered without impact by the attracting or repelling force between the stationary coil and the movable coil of the buffer unit. Accordingly, since the buffer unit is of a non-contact type, there is no mechanical impact and thus no large impact stress, and as it has simple configuration, the reliability is improved and the maintenance can be conducted more easily. (Yoshihara, H.)

  4. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  5. The plutonium product: design of the rod and of the assembly

    International Nuclear Information System (INIS)

    Francillon, G.

    1985-10-01

    On the base of physical and experimental data the aim to be reached is to design a mixed oxide-fuel rod and a mixed oxide-fuel assembly which will be introduced in a PWR type reactor while ensuring the operation and safety of the unit required presently. This paper presents successively the MOX fuel rod and the MOX fuel assembly [fr

  6. Control rod shadowing and anti-shadowing effects in a large gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Girardin, G.; Chawla, R.; Rimpault, G.; Coddington, P.

    2007-01-01

    An investigation of control rod shadowing and anti-shadowing (interaction) effects has been carried out in the context of a design study of the control rod pattern for the large 2400 MWth Generation IV Gas-cooled Fast Reactor (GFR). For the calculations, the deterministic code system ERANOS-2.0 has been used, in association with a full core model including a European Fast Reactor (EFR)-type pattern for the control rods. More specifically, the core contains a total of 33 control (CSD) and safety (DSD) rods implemented in three banks: -1) a first bank of 6 CSD rods, placed at 64 cm from core centre in the inner fuel zone (Pu content 16.3 % vol.), -2) a safety bank consisting of 9 DSD rods, at an average distance of 118 cm, and -3) a third bank with 18 CSD rods, placed at 171 cm, i.e. at the interface between the inner and outer (Pu content 19.2 % vol.) core regions. Each control rod has been modelled as a homogeneous material containing 90%-enriched B 4 C, steel and helium. Considerable shadowing effects have been observed between the first bank and the safety bank, as also between individual rods within the first bank. Large anti-shadowing effects take place in an even greater number of the studied rod configurations. The largest interaction is between the two CSD banks, the anti-shadowing value being 46% in this case, implying that the total rod worth is increased by a factor of almost 2 when compared to the sum of the individual bank values. Additional investigations have been performed, in particular the computation of the first order eigenvalue and the eigenvalue separation. The main finding is that the interactions are lower when one of the control rod banks is located at a radial position corresponding to half the core radius. (authors)

  7. POWER LEVEL EFFECT IN A PWR ROD EJECTION ACCIDENT

    International Nuclear Information System (INIS)

    Diamond, D.J.; Bromley, B.P.; Aronson, A.L.

    2002-01-01

    The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS , a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation

  8. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  9. Key developments of a rod control system - 15101

    International Nuclear Information System (INIS)

    Pouillot, M.; Jegou, H.; Duthou, A.

    2015-01-01

    The aim of the Rod Control System is to carry out the insertion and withdrawal of control rod clusters to provide the power required by the grid (G-mode control), to control the temperature of the reactor, or to provide negative reactivity margin when the reactor is shut down. The rod control system is not classified important for safety, but its correct operation is essential for the availability of the reactor, as the spurious drop of a single cluster usually results in a reactor trip. Rolls-Royce has been designing, manufacturing and providing rod control systems since 1977, in France, China, Belgium, Korea, and South Africa, as an original manufacturer and for modernization projects. All the corresponding nuclear units share the following features, key points for the system design: -) The power source is a three-phased 260 Vac with neutral, provided by zigzag-coupled alternators; -) The Control Rod Drive Mechanisms (CRDM) are 'three-coil type': Stationary Gripper (SG), Movable Gripper (MG) and Lift Coil (LC); -) Rod clusters are arranged in banks and sub-banks, the bank being composed of one or two sub-banks and a sub-bank is a set of 4 clusters moved simultaneously, the central cluster being an exception; and -) Most of those reactors are operated in G-mode (load following). (authors)

  10. Study on anti-seismic test of control rod driving system suspended by magnetic force

    International Nuclear Information System (INIS)

    Zhang Zhihua; Qian Dazhi; Xu Xianqi; Huang Hongwen; Zhang Zhengming; Wu Xinxin; Hu Xiao

    2012-01-01

    To verify the stability, reliability and security function in extreme conditions, the anti-seismic test of control rod drive line was conducted. Drop-time of control rod drive line in different earthquake intensities was got. The response and strain values of control rod drive line acceleration on SL-1, SL-2 level were measured. Safety functions of control rod drive line were validated in different work conditions. Anti-seismic test data shows that the driving system can keep the structure's integrality and realize operation function under OBE and SSE. (authors)

  11. Preliminary performance test of control rod position indicator for ballscrew type CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hu, H.; Lee, J. S.; Kim, J. I.

    2003-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The prototype of control rod position indicator having the high performance for the ballscrew type CEDM was developed on the basis of RSPT technology identified through the survey. The characteristics of control rod position indicator was defined and documented through design procedure and preliminary performance test

  12. Control rod position detection device

    International Nuclear Information System (INIS)

    Akita, Haruo; Ogiwara, Sakae.

    1996-01-01

    The device of the present invention is used in a back-up shut down system of an LMFBR type reactor which is easy for maintenance, has high reliability and can recognize the position of control rods accurately. Namely, a permanent magnet is disposed to a control rod extension tube connected to the lower portion of the control rod. The detector guide tube is disposed in the vicinity of the control rod extension tube. A detector having a detection coil is inserted into a detector tube. With such constitution, the control rod can be detected at one position using the following method. (1) the movement of the magnetic field of the permanent magnet is detected by the detection coil. (2) a plurality of grooves are formed on the control rod extension tube, and the movement of the grooves is detected. In addition, the detection coil is inserted into the detector guide tube, and the signals from the detection coil are inputted to a signal processing circuit disposed at the outside of the reactor vessel using an MI cable to enable the maintenance of the detector. Further, if the detector comprises a detection coil and an excitation coil, the position of a dropped control rod can be recognized at a plurality of points. (I.S.)

  13. Control rod position control device

    International Nuclear Information System (INIS)

    Ubukata, Shinji.

    1997-01-01

    The present invention provides a control rod position control device which stores data such as of position signals and driving control rod instruction before and after occurrence of abnormality in control for the control rod position for controlling reactor power and utilized the data effectively for investigating the cause of abnormality. Namely, a plurality of individual control devices have an operation mismatching detection circuit for outputting signals when difference is caused between a driving instruction given to the control rod position control device and the control rod driving means and signals from a detection means for detecting an actual moving amount. A general control device collectively controls the individual control devices. In addition, there is also disposed a position storing circuit for storing position signals at least before and after the occurrence of the control rod operation mismatching. With such procedures, the cause of the abnormality can be determined based on the position signals before and after the occurrence of control rod mismatching operation stored in the position storing circuit. Accordingly, the abnormality cause can be determined to conduct restoration in an early stage. (I.S.)

  14. Status of rod consolidation, 1988

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs

  15. Inspecting method for fuel rods

    International Nuclear Information System (INIS)

    Watanabe, Masaaki; Kogure, Sumio.

    1976-01-01

    Purpose: To precisely detect the response of flaw in clad tube and submerged fuel pellets from a relationship between the surface of fuel rod and internal signal. Constitution: Ultrasonic reflected waves from the surface of fuel rods and the interior are detected and either one of fuel rod or ultrasonic flaw detecting contact is rotated to thereby precisely detect the response of the flaw of clad tube and submerged fuel pellets from a relationship between said surface and the interior. It will be noted that the ultrasonic flaw detecting contact used is of the line-focus type, the incident angle of ultrasonic wave from the ultrasonic flaw detecting contact relative to the fuel rod is the angle of skew, that is, the ultrasonic flaw detecting contact is not perpendicular to a center axis of the fuel rod but is slightly displace. That is, the use of the aforesaid contact may facilitate discrimination between the surface flaw of the fuel rod and the response of submergence, and in addition, the employment of the aforesaid incident angle makes it hard to receive reflected waves from the surface of the fuel rod which is great in terms of energy to facilitate discrimination of waves responsive to submergence. (Kawakami, Y.)

  16. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  17. Spacers for fuel rod clusters

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    The proposition deals with the fixing of nuclear fuel element rods in a grid which consists of a number of crossed Zy-plates which form cells. The rectangular cells have projections which serve as spacers for the fuel rods. According to the invention there are additional butt straps which can be moved in such a way that insertion and extraction of the fuel rods can be done without obstruction and they can be spring-loaded hold in their final position. (UWI) [de

  18. Drilling Experiments of Dummy Fuel Rods Using a Mock-up Drilling Device and Detail Design of Device for Drilling of Irradiated Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yong; Lee, H. K.; Chun, Y. B.; Park, S. J.; Kim, B. G

    2007-07-15

    KAERI are developing the safety evaluation method and the analysis technology for high burn-up nuclear fuel rod that is the project, re-irradiation for re-instrumented fuel rod. That project includes insertion of a thermocouple in the center hole of PWR nuclear fuel rod with standard burn-up, 3,500{approx}4,000MWD/tU and then inspection of the nuclear fuel rod's heat performance during re-irradiation. To re-fabricate fuel rod, two devices are needed such as a drilling machine and a welding machine. The drilling machine performs grinding a center hole, 2.5 mm in diameter and 50 mm in depth, for inserting a thermocouple. And the welding machine is used to fasten a end plug on a fuel rod. Because these two equipment handle irradiated fuel rods, they are operated in hot cell blocked radioactive rays. Before inserting any device into hot cell, many tests with that machine have to be conducted. This report shows preliminary experiments for drilling a center hole on dummy of fuel rods and optimized drilling parameters to lessen operation time and damage of diamond dills. And the design method of a drilling machine for irradiated nuclear fuel rods and detail design drawings are attached.

  19. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  20. Control rod drive

    International Nuclear Information System (INIS)

    Kojima, Akira.

    1989-01-01

    In the control rod drive for a BWR type reactor, etc., according to this invention, the lower limit flow rate is set so as to keep the restriction for stability upon spectral shift operation. The setting condition for keeping the restriction is the lowest pump speed and the lower limit for the automatic control of the flow rate, which are considered to be important in view of the stablility from the actual power state. In view of the above, it is possible to keep the reactor core stably even in a case where such a transient phenomenon occurs that the recycling flow rate has to be run back to the lowest pump speed during spectral shift opeeration or in a case where the load demand is reduced and the flow rate is decreased by an automatic mode as in night operation. Accordingly, in the case of conducting the spectral shift operation according to this invention, the operation region capable of keeping the reactor core state stably during operation can be extended. (I.S.)

  1. Control rod drive

    International Nuclear Information System (INIS)

    Watando, Kosaku; Tanaka, Yuzo; Mizumura, Yasuhiro; Hosono, Kazuya.

    1975-01-01

    Object: To provide a simple and compact construction of an apparatus for driving a drive shaft inside with a magnetic force from the outside of the primary system water side. Structure: The weight of a plunger provided with an attraction plate is supported by a plunger lift spring means so as to provide a buffer action at the time of momentary movement while also permitting the load on lift coil to be constituted solely by the load on the drive shaft. In addition, by arranging the attraction plate and lift coil so that they face each other with a small gap there-between, it is made possible to reduce the size and permit efficient utilization of the attracting force. Because of the small size, cooling can be simply carried out. Further, since there is no mechanical penetration portion, there is no possibility of leakage of the primary system water. Furthermore, concentration of load on a latch pin is prevented by arranging so that with a structure the load of the control rod to be directly beared through the scrum latch. (Kamimura, M.)

  2. Generation of heat on fuel rod in cosine pattern by using induction heating

    International Nuclear Information System (INIS)

    Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

    2017-01-01

    Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

  3. Means for driving control rod

    International Nuclear Information System (INIS)

    Sato, Haruo; Sasaki, Masayoshi.

    1974-01-01

    Object: To enable wire rope to be readily removed from guide pulleys for the inspection or replacement of control rods. Structure: A pair of guide pulleys disposed to oppose each other are provided on their periphery with respective notches which are arranged in a staggered fashion. In this way, the rope is made to be removed from the notches for inspection of the control rod or for other purposes. (Kamimura, M.)

  4. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

    This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

  5. Construction and testing of a 'fork' type central rod

    International Nuclear Information System (INIS)

    Guicherd, R.; Meunier, C.

    1964-01-01

    Aim of the control rod: - Reduction of the flux peaks - Increase of the efficiency - simplification of the mechanics - without modification of the grid nor of the standard fuel plates. Results of the tests: - Reduction of the flux peaks by 29 per cent - Increase of the efficiency by 20 per cent - The suppleness of the plates and the decrease in the number of parts results in greater safety during operation. Conclusions: Study of a generalized adaptation of this type of rod which presents definite advantages. (authors) [fr

  6. Fabrication of internally instrumented reactor fuel rods

    International Nuclear Information System (INIS)

    Schmutz, J.D.; Meservey, R.H.

    1975-01-01

    Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

  7. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  8. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    Science.gov (United States)

    Oakes, L.C.; Walker, C.S.

    1959-12-15

    ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

  9. The thermo-mechanics of the PWR fuel rod

    International Nuclear Information System (INIS)

    Barral, J.C.; Gautier, B.; Chaigne, G.

    1999-01-01

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  10. 3-D rod ejection analysis using a conservative methodology

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min Ho; Park, Jin Woo; Park, Guen Tae; Um, Kil Sup; Ryu, Seok Hee; Lee, Jae Il; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The point kinetics model which simplifies the core phenomena and physical specifications is used for the conventional rod ejection accident analysis. The point kinetics model is convenient to assume conservative core parameters but this simplification loses large amount of safety margin. The CHASER system couples the three-dimensional core neutron kinetics code ASTRA, the sub-channel analysis code THALES and the fuel performance analysis code FROST. The validation study for the CHASER system is addressed using the NEACRP three-dimensional PWR core transient benchmark problem. A series of conservative rod ejection analyses for the APR1400 type plant is performed for both hot full power (HFP) and hot zero power (HZP) conditions to determine the most limiting cases. The conservative rod ejection analysis methodology is designed to properly consider important phenomena and physical parameters.

  11. Radioactive lightning rods waste treatment

    International Nuclear Information System (INIS)

    Vicente, Roberto; Dellamano, Jose C.; Hiromoto, Goro

    2008-01-01

    Full text: In this paper, we present alternative processes that could be adopted for the management of radioactive waste that arises from the replacement of lightning rods with attached Americium-241 sources. Lightning protectors, with Americium-241 sources attached to the air terminals, were manufactured in Brazil until 1989, when the regulatory authority overthrew the license for fabrication, commerce, and installation of radioactive lightning rods. It is estimated that, during the license period, about 75,000 such devices were set up in public, commercial and industrial buildings, including houses and schools. However, the policy of CNEN in regard to the replacement of the installed radioactive rods, has been to leave the decision to municipal governments under local building regulations, requiring only that the replaced rods be sent immediately to one of its research institutes to be treated as radioactive waste. As a consequence, the program of replacement proceeds in a low pace and until now only about twenty thousand rods have reached the waste treatment facilities The process of management that was adopted is based primarily on the assumption that the Am-241 sources will be disposed of as radioactive sealed sources, probably in a deep borehole repository. The process can be described broadly by the following steps: a) Receive and put the lightning rods in initial storage; b) Disassemble the rods and pull out the sources; c) Decontaminate and release the metal parts to metal recycling; d) Store the sources in intermediate storage; e) Package the sources in final disposal packages; and f) Send the sources for final disposal. Up to now, the disassembled devices gave rise to about 90,000 sources which are kept in storage while the design of the final disposal package is in progress. (author)

  12. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

  13. An evaluation of control rod motion simulator of research reactor

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

  14. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator has been carried out testing of the reactor control rod using a servomotor. Reactor control rod motion at any point should be in the right position, one of the motors that can move in a precise and correct the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servomotor function test for motor work to ensure the performance of the appliance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage Vout nets at 24 V, 6.5 A with 12 Ω load deviation obtained V0 = V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125%, next to the breakdown voltage Vout nets at 12V, 4.2 A with a 6 Ω load deviation obtained V0 = V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on Vout 24 V, 4.5 A with 12 Ω load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%. (author)

  15. The Third ATLAS ROD Workshop

    CERN Multimedia

    Poggioli, L.

    A new-style Workshop After two successful ATLAS ROD Workshops dedicated to the ROD hardware and held at the Geneva University in 1998 and in 2000, a new style Workshop took place at LAPP in Annecy on November 14-15, 2002. This time the Workshop was fully dedicated to the ROD-TDAQ integration and software in view of the near future integration activities of the final RODs for the detector assembly and commissioning. More precisely, the aim of this workshop was to get from the sub-detectors the parameters needed for T-DAQ, as well as status and plans from ROD builders. On the other hand, what was decided and assumed had to be stated (like EB decisions and URDs), and also support plans. The Workshop gathered about 70 participants from all ATLAS sub-detectors and the T-DAQ community. The quite dense agenda allowed nevertheless for many lively discussions, and for a dinner in the old town of Annecy. The Sessions The Workshop was organized in five main sessions: Assumptions and recommendations Sub-de...

  16. Future plans for performance analysis and maintenance/inspection optimization of shutoff rods based on the case study of Bruce Power Unit-3 Shutoff Rod 5 inspection

    International Nuclear Information System (INIS)

    Nasimi, E.; Gabbar, H.A.

    2011-01-01

    Shutdown System 1 (SDS1) is a preferred method for a quick shutdown of nuclear fission process in CANDU (CANada Deuterium Uranium) reactor units. Failure of a routine SDS1 safety test during Fall 2009 outage resulted in the need to develop and execute a new methodology for Shutoff Rod inspection and re-evaluate the known degradation mechanisms and failure modes. This paper describes the development of this methodology and the obtained results. It also proposes several alternative solutions for the future performance analysis and maintenance/inspection optimization for SDS1 Shutoff Rods based on the Bruce Power Unit-3 Shutoff Rod 5 case study. (author)

  17. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  18. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  19. Dynamic insertion analysis of control rods of BWR under seismic excitation

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Koide, Yuichi; Fukushi, Naoki; Ishigaki, Hirokuni; Okumura, Kazue

    2007-01-01

    The dynamic insertion characteristics of the control rods for the boiling water reactors under the seismic excitation are investigated using non-linear analytical models. The control rod insertion capability is one of the most important items for the safety of nuclear power plants under the seismic events. Predicting the control rod insertion behavior during the earthquake is important in the course of the control rod seismic design. We developed the analytical models using the finite element method (FEM). The effect of the interaction force between the control rod and the fuel assemblies is considered in the non-linear analysis. This interaction force courses the resistance force to the control rod during its insertion behavior. The validity of analytical methods was confirmed by comparing the analytical results with the experimental ones. Using the analytical models, the effects of input seismic motion and structural parameters of the control rods and the fuel assemblies, such as the thickness of the channel box, on the insertion time are investigated. These analytical methods can predict insertion time of the control rod, and are useful for the seismic design of the control rod assemblies. (author)

  20. Control rod drive for vertical movement

    International Nuclear Information System (INIS)

    Suskov, I.I.; Gorjunov, V.S.; Zajcev, B.I.; Derevjankin, N.E.; Petrov, V.A.; Istomin, S.D.; Kovalencik, D.I.; Archipov, E.A.; Serebrjakov, V.I.; Kacalin, V.S.

    1982-01-01

    The control of the rod repositioning gear unit and the control unit of the profile grab of the control rod drive for the alkali metal-cooled fast breeder reactor is achieved by an electromotor being arranged outside the hermetic drive casing. The guide tube is directly repositioned by the rod repositioning gear unit. Coupling control of the drive with the control rod is done in the lower operative position of the control rod and that because of the interaction of the tie rod arranged on the spring-mounted control rod with the induction transmitter for the lower position of the control rod. In the transfer position the rod is fixed within the guide tube. (orig.)

  1. Maximum/minimum asymmetric rod detection

    International Nuclear Information System (INIS)

    Huston, J.T.

    1990-01-01

    This patent describes a system for determining the relative position of each control rod within a control rod group in a nuclear reactor. The control rod group having at least three control rods therein. It comprises: means for producing a signal representative of a position of each control rod within the control rod group in the nuclear reactor; means for establishing a signal representative of the highest position of a control rod in the control rod group in the nuclear reactor; means for establishing a signal representative of the lowest position of a control rod in the control rod group in the nuclear reactor; means for determining a difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; means for establishing a predetermined limit for the difference between the signal representative of the position of the highest control rod and the signal representative of the position of the lowest control rod; and means for comparing the difference between the signals with the predetermined limit. The comparing means producing an output signal when the difference between the signals exceeds the predetermined limit

  2. Refabricated and instrumented fuel rods

    International Nuclear Information System (INIS)

    Silberstein, K.

    2005-01-01

    Nuclear Fuel for power reactors capabilities evaluation is strongly based on the intimate knowledge of its behaviour under irradiation. This knowledge can be acquired from refabricated and instrumented fuel rods irradiated at different levels in commercial reactors. This paper presents the development and qualification of a new technique called RECTO related to a double-instrumented rod re-fabrication process developed by CEA/LECA hot laboratory facility at CADARACHE. The technique development includes manufacturing of the properly dimensioned cavity in the fuel pellet stack to house the thermocouple and the use of a newly designed pressure transducer. An analytic irradiation of such a double-instrumented fuel rod will be performed in OSIRIS test reactor starting October 2004. (Author)

  3. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 2 discusses the following topics: Fuel Rod Extraction System Test Results and Analysis Reports and Clamping Table Test Results and Analysis Reports

  4. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 1 of Volume IV, discusses: Process overview functional descriptions; Control system descriptions; Support system descriptions; Maintenance system descriptions; and Process equipment descriptions

  5. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  6. Advanced gray rod control assembly

    Science.gov (United States)

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  7. Lifting device for drilling rods

    Energy Technology Data Exchange (ETDEWEB)

    Radzivilovich, L L; Laptev, A G; Lipkovich, V A

    1982-01-01

    A lifter is proposed for drilling rods including a spacer stand with rotating bracket, boom with by-pass rollers, spacing and lifting hydrocylinders with rods and flexible tie mechanism. In order to improve labor productivity by improving maneuverability and to increase the maintenance zone, the lifter is equipped with a hydrocylinder of advance and a cross piece which is installed with the possibility of forward and rotational movement on the stand, and in which by means of the hydrocylinder of advance a boom is attached. Within the indicated boom there is a branch of the flexible tie mechanism with end attached with the possibility of regulation over the length on a rotating bracket, while the rod of the lifting hydrocylinder is connected to the cross piece.

  8. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase III of the Prototypical Rod Consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod Consolidation System as described in the NUS Phase II Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase III effort the system was tested on a component, subsystem, and system level. Volume IV provides the Operating and Maintenance Manual for the Prototypical Rod Consolidation System that was installed at the Cold Test Facility. This document, Book 4 of Volume IV, discusses: Off-normal operating and recovery procedures; Emergency response procedures; Troubleshooting procedures; and Preventive maintenance procedures

  9. Determination of the control rod worth for research reactors

    International Nuclear Information System (INIS)

    Aldama, D.L.; Gual, M.R.

    2000-01-01

    Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

  10. A probabilistic design method for LMFBR fuel rods

    International Nuclear Information System (INIS)

    Peck, S.O.; Lovejoy, W.S.

    1977-01-01

    Fuel rod performance analyses for design purposes are dependent upon material properties, dimensions, and loads that are statistical in nature. Conventional design practice accounts for the uncertainties in relevant parameters by designing to a 'safety factor', set so as to assure safe operation. Arbitrary assignment of these safety factors, based upon a number of 'worst case' assumptions, may result in costly over-design. Probabilistic design methods provide a systematic way to reflect the uncertainties in design parameters. PECS-III is a computer code which employs Monte Carlo techniques to generate the probability density and distribution functions for time-to-failure and cumulative damage for sealed plenum LMFBR fuel rods on a single rod or whole core basis. In Monte Carlo analyses, a deterministic model (that maps single-valued inputs into single-valued outputs) is coupled to a statistical 'driver'. Uncertainties in the input are reflected by assigning probability densities to the input parameters. Dependent input variables are considered multivariate normal. Independent input variables may be arbitrarily distributed. Sample values are drawn from these input densities, and a complete analysis is done by the deterministic model to generate a sample point in the output distribution. This process is repeated many times, and the number of times each output value occurs is accumulated. The probability that some measure of rod performance will fall within given limits is estimated by the relative frequency with which the Monte Carlo samples fall within tho

  11. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  12. Control rod housing alignment apparatus

    International Nuclear Information System (INIS)

    Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

    1991-01-01

    This paper discusses an alignment device for precisely locating the position of the top of a control rod drive housing from an overlying and corresponding hole and alignment pin in a core plate within a boiling water nuclear reactor. It includes a shaft, the shaft having a length sufficient to extend from the vicinity of the top of the control rod drive housing up to and through the hole in the core plate; means for registering the top of the shaft to the hole in the core plate, the registering means including means for registering with an alignment pin in the core plate adjacent the hole

  13. Control rod guide tube assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A nuclear fuel assembly including sleeves telescoped over end portions of control rod guide tubes which bear against internal shoulders of the sleeves. Upper ends of the sleeves protrude beyond a control rod guide tube spider and are locked in place by means of a resilient cellular lattice or lock that is seated in mating grooves in the outer surfaces of the sleeves. A grapple is provided for disengaging the entire lock structure spider and associated washers, springs and a grill from the end of the fuel assembly in order to enable these components to be removed and subsequently replaced on the fuel assembly after inspection and repair. (UK)

  14. Resolution No. 58/03 Concerning the use of radioactive lightning rods

    International Nuclear Information System (INIS)

    2003-01-01

    This resolution provides that the importation and purchase of radioactive lightning rods throughout the national territory is prohibited also not the assembly of new radioactive lightning rod is allowed although these have in storage. All currently employs radioactive lightning rods should replace conventional, within ten (10) years from the entry into force of this Resolution. The radioactive lightning rods in use, for any reason lose tightness sources must be removed immediately. In all cases should proceed to the safe management of radioactive lightning rod as disused source with the Center for Health and Safety of Radiation or other entity authorized to do so, in order to avoid undue risk to lives, property and the environment environment.

  15. Flow resistance in rod assemblies

    International Nuclear Information System (INIS)

    Korsun, A.S.; Sokolova, M.S.

    2000-01-01

    The general form of relation between the resistance force and the velocity vector, resistance tensor structure and possible types of anisotropy in the flow thorough such structures as rod or tube assemblies are under discussion. Some questions of experimental determination of volumetric resistance force tensor are also under consideration. (author)

  16. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  17. Control rod driving hydraulic device

    International Nuclear Information System (INIS)

    Sugano, Hiroshi.

    1993-01-01

    In a control rod driving hydraulic device for an improved BWR type reactor, a bypass pipeline is disposed being branched from a scram pipeline, and a control orifice and a throttle valve are interposed to the bypass pipeline for restricting pressure. Upon occurrence of scram, about 1/2 of water quantity flowing from an accumulator of a hydraulic control unit to the lower surface of a piston of control rod drives by way of a scram pipeline is controlled by the restricting orifice and the throttle valve, by which the water is discharged to a pump suction pipeline or other pipelines by way of the bypass pipeline. With such procedures, a function capable of simultaneously conducting scram for two control rod drives can be attained by one hydraulic control unit. Further, an excessive peak pressure generated by a water hammer phenomenon in the scram pipeline or the control rod drives upon occurrence of scram can be reduced. Deformation and failure due to the excessive peak pressure can be prevented, as well as vibrations and degradation of performance of relevant portions can be prevented. (N.H.)

  18. Evolution of fuel rod support under irradiation impact on the mechanical behaviour of fuel assemblies

    International Nuclear Information System (INIS)

    Billerey, Antoine; Waeckel, Nicolas

    2005-01-01

    New fuel management targets imply to increase fuel assembly discharge burnup. Therefore, the prediction of the mechanical behaviour of the irradiated fuel assembly is essential such as excessive fuel assembly distortion induce incomplete Rod Cluster Control Assembly insertion problems (safety issue) or fuel rod vibration induced wear leading to leaking rods (plant operation problems). Within this framework, one of the most important parameter is the knowledge of the fuel rod support in the grid cell because it directly governs the mechanical behaviour of the fuel assembly and consequently allows to predict the behaviour of irradiated structures in terms of (1) axial and lateral deformation (global behaviour of the assembly) and (2) rod vibration induced wear (local behaviour of the rod). Generally, fuel rod support is provided by a spring-dimple system fixed to the grid. During irradiation, the spring force decreases and a gap between the rod and the spring may occur. This phenomenon is due to (1) stress relieving in the spring and in the dimples, (2) grid growth and (3) reduction of the rod diameter. Two models have been developed to predict the behaviour of the rod in the cell. The first model is dedicated to the evaluation of the spring force relaxation during irradiation. The second one can assess the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (1) the creep laws of the grid materials, (2) the growth law of the grid, (3) the evolution of rod diameter and (4) the design of the fuel rod support. The aim of this paper is to: (1) evaluate the consequences of grid support design modifications on the rod vibration sensitivity in terms of predicted rod to grid maximum gap during irradiation and time in operation with an open rod to grid gap, (2) evaluate, using a linear or non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the overall mechanical

  19. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  20. Control rods in LMFBRs: a physics assessment

    International Nuclear Information System (INIS)

    McFarlane, H.F.; Collins, P.J.

    1982-08-01

    This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B 4 C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined

  1. Experiments on the quench behavior of fuel rods

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-01-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  2. Experiments on the quench behavior of fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.

    1995-08-01

    Because of the importance of the observed reflood phenomena for safety of current and future LWRs, the Forschungszentrum Karlsruhe (FZKA) started a program to investigate the mechanisms of quench-induced oxidation of Zircaloy. A small scale test-rig was designed and built in which it is possible to quench single Zircaloy rods by water and steam. The report describes the status of this work in May 1995. Some experimental results are presented. (orig./HP)

  3. Stability and failure analysis of steering tie-rod

    Science.gov (United States)

    Jiang, GongFeng; Zhang, YiLiang; Xu, XueDong; Ding, DaWei

    2008-11-01

    A new car in operation of only 8,000 km, because of malfunction, resulting in lost control and rammed into the edge of the road, and then the basic vehicle scrapped. According to the investigation of the site, it was found that the tie-rod of the car had been broken. For the subjective analysis of the accident and identifying the true causes of rupture of the tierod, a series of studies, from the angle of theory to experiment on the bended broken tie-rod, were conducted. The mechanical model was established; the stability of the defective tie-rod was simulated based on ANSYS software. Meanwhile, the process of the accident was simulated considering the effect of destabilization of different vehicle speed and direction of the impact. Simultaneously, macro graphic test, chemical composition analysis, microstructure analysis and SEM analysis of the fracture were implemented. The results showed that: 1) the toughness of the tie-rod is at a normal level, but there is some previous flaws. One quarter of the fracture surface has been cracked before the accident. However, there is no relationship between the flaw and this incident. The direct cause is the dynamic instability leading to the large deformation of impact loading. 2) The declining safety factor of the tie-rod greatly due to the previous flaws; the result of numerical simulation shows that previous flaw is the vital factor of structure instability, on the basis of the comparison of critical loads of the accident tie-rod and normal. The critical load can decrease by 51.3% when the initial defect increases 19.54% on the cross-sectional area, which meets the Theory of Koiter.

  4. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  5. Design and Experimental Evaluation of an Electromagnetic Acceleration System for Fast Safety Rods; Etude Theorique et Experimentale d'un Mecanisme Electromagnetique d'acceleration pour des Barres de Securite; Proektirovanie i ehksperimental'naya otsenka ehlektromagnitnoj sistemy uskoreniya dlya avarijnykh sterzhnej bystrogo dejstviya; Proyecto y Estudio Experimental de un Sistema Electromagnetico de Aceleracion para Barras Rapidas de Seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Dosch, P.; Kraus, H. -J.; Uhrig, H. [Kernreaktor Bau- und Betriebs-Gesellschaft m.b.H., Karlsruhe, Federal Republic of Germany (Germany)

    1964-06-15

    The safety system of a reactor must have a very short delay time if it is to control very fast reactivity disturbances. The electronic parts.of a safety system may be equipped with short response times very easily, whereas reduction of the mechanical release time of the safety rod is a difficult task More exact investigations show that, besides the short release time, high initial speed of the rod would be an advantage. At equal total times of insertion a rod having uniform velocity is by far superior to a rod having uniform acceleration. From this it is seen that a pulse-shaped acceleration is the proper way of driving safety rods to meet fast reactivity disturbances, because such acceleration takes care of short delay times and provides high initial velocities. An electromagnetic pulse generator was found to be suitable for application in safety rods. It features these advantages: (a) The components are few and of simple and sturdy design, which may easily be included in a safety rod and can also be designed for conditions near the core (high levels of temperature and radiation); (b) Driving energy is stored in an electric capacitor and thus is also available in case the power supply fails; and (c) The design can easily be made in such a way. as to permit an unrestricted free fall in case the pulse fails. The results of experimental investigations conducted in electromagnetic pulse generators are described. It is possible to accelerate a safety rod after less than a millisecond to initial speeds of several.metres per second, depending upon the weight of the rod, of course. The design of a fast safety rod for a fast zero-energy assembly (SNEAK) is described. A prototype of this safety rod is being tested. The results have thus far been very satisfactory. (author) [French] Les barres de securite d'un reacteur doivent pouvoir etre abaissees dans des delais tres courts si l'on veut compenser des variations tres brusques de reactivite. Il est tres facile de munir la

  6. Process and apparatus for controlling control rods

    International Nuclear Information System (INIS)

    Gebelin, B.; Couture, R.

    1987-01-01

    This process and apparatus is characterized by 2 methods, for examination of cluster of nuclear control rods. Foucault current analyzer which examines fraction by fraction all the control rods. This examination is made by rotation of the cluster. Doubtful rods are then analysed by ultrasonic probe [fr

  7. Fuel followed control rod installation at AFRRI

    International Nuclear Information System (INIS)

    Moore, Mark; Owens, Chris; Forsbacka, Matt

    1992-01-01

    Fuel Followed Control Rods (FFCRs) were installed at the Armed Forces Radiobiology Research Institute's 1 MW TRIGA Reactor. The procedures for obtaining, shipping, and installing the FFCRs is described. As part of the FFCR installation, the transient rod drive was relocated. Core performance due to the addition of the fuel followed control rods is discussed. (author)

  8. Solitary waves on nonlinear elastic rods. II

    DEFF Research Database (Denmark)

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

    1987-01-01

    In continuation of an earlier study of propagation of solitary waves on nonlinear elastic rods, numerical investigations of blowup, reflection, and fission at continuous and discontinuous variation of the cross section for the rod and reflection at the end of the rod are presented. The results ar...... are compared with predictions of conservation theorems for energy and momentum....

  9. ELECTRIC FIELD MEASUREMENT IN ROD-DISCONTINUED ...

    African Journals Online (AJOL)

    2014-06-30

    Jun 30, 2014 ... the electrogeometrical model using a laboratory experimental rod-plane air gap arrangement with a lightning conductor (Franklin rod or horizontal conductor). The stepped leader could be represented by the rod electrode under a negative lightning impulse voltage having a level leading to breakdown with ...

  10. Testing device for control rod drives

    International Nuclear Information System (INIS)

    Hayakawa, Toshifumi.

    1992-01-01

    A testing device for control rod drives comprises a logic measuring means for measuring an output signal from a control rod drive logic generation circuit, a control means for judging the operation state of a control rod and a man machine interface means for outputting the result of the judgement. A driving instruction outputted from the control rod operation device is always monitored by the control means, and if the operation instruction is stopped, a testing signal is outputted to the control rod control device to simulate a control rod operation. In this case, the output signal of the control rod drive logic generation circuit is held in a control rod drive memory means and intaken into a logic analysis means for measurement and an abnormality is judged by the control means. The stopping of the control rod drive instruction is monitored and the operation abnormality of the control rod is judged, to mitigate the burden of an operator. Further, the operation of the control rod drive logic generation circuit can be confirmed even during a nuclear plant operation by holding the control rod drive instruction thereby enabling to improve maintenance efficiency. (N.H.)

  11. 21 CFR 876.4270 - Colostomy rod.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Colostomy rod. 876.4270 Section 876.4270 Food and... GASTROENTEROLOGY-UROLOGY DEVICES Surgical Devices § 876.4270 Colostomy rod. (a) Identification. A colostomy rod is a device used during the loop colostomy procedure. A loop of colon is surgically brought out through...

  12. Spider and burnable poison rod combinations

    International Nuclear Information System (INIS)

    Edwards, G.T.; Schluderberg, D.C.

    1980-01-01

    An improved design of burnable poison rods and associated spiders used in fuel assemblies of pressurized water power reactor cores, is described. The rods are joined to the spider arms in a manner which is proof against the reactor core environment and yet allows the removal of the rods from the spider simply, swiftly and delicately. (U.K.)

  13. Self-Assembly of Rod-Coil Block Copolymers

    National Research Council Canada - National Science Library

    Jenekhe, S

    1999-01-01

    ... the self-assembly of new rod-coil diblock, rod- coil-rod triblock, and coil-rod-coil triblock copolymers from solution and the resulting discrete and periodic mesostmctares with sizes in the 100...

  14. Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

    International Nuclear Information System (INIS)

    Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

    2001-09-01

    Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

  15. Development and assessment of modular models of calculation for the interpretation of rod-melting experiments

    International Nuclear Information System (INIS)

    Tuerk, W.

    1980-01-01

    By the example of recalculations of rod-melting experiment it is shown how a modular simulation model for complex systems can be formulated within the scope of RSYST1. The procedure of code development as well as the physical and numerical methods and approximations of the simulation model are described. To each important physical process a code module is assigned. The individual moduls describe heat production, rod heat-up, rod oxidation, rod environment, rod deformation by thermal expansion and can buckling, melting of the rod, rod failure, and flowing off of the melted mass. A comparison of the results for the overall model with the result of different experiments indicates that the phenomena during heat-up and melting of the rod are treated in agreement with the experiments. The results of the calculation model and its submodels are thus largely supported by experiments. Therefore further predictions with a high level of confidence can be made with the model within the scope of reactor safety research. (orig.) [de

  16. Temperature and Stresses Estimation in Reactivity Control Rods for CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Markiewicz, Mario; Estevez, Esteban

    2000-01-01

    The reactivity control rods are a critical component regarding safety.Its correct operation when required must be ensured.For this purpose, this component must maintain its operating capacity during all its residence time and under any foreseen operation condition.To evaluate the behaviour of reactivity control rods, it is necessary to analyse the demands they are exposed to, determining from the mechanical point of view, the residence time in the reactor core.In this report, using analytical calculations, the parameters affecting the performance of the reactivity control rods are analysed, with the objective of determine from the mechanical point of view, its behaviour and residence time

  17. Rolls-Royce digital Rod Control System

    International Nuclear Information System (INIS)

    Pouillot, M.

    2010-01-01

    -graph representing min-max monitoring thresholds, and grouped faults. Plugs allow output current checking per coil. - From the cycler, the output current can be forced on a rack for test, while ensuring that rods are held by another available gripper rack. - Low power dissipation: no need for multiplexing or LC-inhibition thyristors - Commissioning: 3-thyristor well-tried half bridges (no energy storage), fuses for protection of feed-through penetrations. The paper will present in details the technical features of the Rolls-Royce Rod Control System and the main benefits to use it. About Rolls-Royce Rolls-Royce is a global business providing integrated power systems for use on land, at sea and in the air. The Group has a balanced business portfolio with leading market positions. Rolls-Royce has a broad range of civil nuclear expertise, including work related to licensing and safety reviews, engineering design, supply chain management, manufacturing, installation and commissioning of the nuclear island systems and equipments, as well as operational management and through life support. Rolls-Royce benefits of a strong reputation worldwide and of a unique experience in the nuclear market. (author)

  18. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  19. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    Matthews, J.C.

    1978-01-01

    A snubber cartridge assembly is mounted to the nozzle of a control rod drive mechanism to insure that the snubber assembly will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllably exhaust liquid therefrom during a ''scram'' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe ''scram'' of the control rod into the reactor

  20. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J. L.; Howell, C. A.; Smith, J. H.; Vining, G. E.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  1. Individual nuclear fuel rod weighing system

    International Nuclear Information System (INIS)

    Fogg, J.L.; Smith, J.H.; Vining, G.E.; Howell, C.A.

    1985-01-01

    An individual nuclear fuel rod weighing system for rods carried on a tray which moves along a materials handling conveyor is discussed. At a first tray position on the conveyor, a lifting device raises the rods off the tray and places them on an overhead ramp. A loading mechanism conveys the rods singly from the overhead ramp onto an overhead scale for individual weighing. When the tray is at a second position on the conveyor, a transfer apparatus transports each weighed rod from the scale back onto the tray

  2. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nyland, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1990-01-01

    This patent describes an apparatus for loading fuel rods in a desired pattern. It comprises: a carousel having a plurality of movable gondolas for stocking thereon fuel rods of known enrichments; an elongated magazine defining a matrix of elongated slots being open at their forward ends for receiving fuel rods; a workstation defining a fuel rod feed path; and a holder and indexing mechanism for movably supporting the magazine and being actuatable for moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  3. ELECTROMAGNETIC APPARATUS FOR MOVING A ROD

    Science.gov (United States)

    Young, J.N.

    1958-04-22

    An electromagnetic apparatus for moving a rod-like member in small steps in either direction is described. The invention has particular application in the reactor field where the reactor control rods must be moved only a small distance and where the use of mechanical couplings is impractical due to the high- pressure seals required. A neutron-absorbing rod is mounted in a housing with gripping uaits that engage the rod, and coils for magnetizing the gripping units to make them grip, shift, and release the rod are located outside the housing.

  4. Snubber assembly for a control rod drive

    International Nuclear Information System (INIS)

    1976-01-01

    A snubber cartridge assembly is described which is mounted to the nozzle of a control rod drive mechanism to insure that it will be located within the liquid filled section of a nuclear reactor vessel whenever the control rod drive is assembled thereto. The snubber assembly includes a piston-mounted proximate to the control rod connecting end of the control rod drive leadscrew to allow the piston to travel within the liquid filled snubber cartridge and controllable exhaust the liquid during a 'scram' condition. The snubber cartridge provides three separate areas of increasing resistance to piston travel to insure a speedy but safe 'scram' of the control rod into the reactor

  5. Rod cluster having improved vane configuration

    International Nuclear Information System (INIS)

    Shockling, L.A.; Francis, T.A.

    1989-01-01

    This patent describes a pressurized water reactor vessel, the vessel defining a predetermined axial direction of the flow of coolant therewithin and having plural spider assemblies supporting, for vertical movement within the vessel, respective clusters of rods in spaced, parallel axial relationship, parallel to the predetermined axial direction of coolant flow, and a rod guide for each spider assembly and respective cluster of rods. The rod guide having horizontally oriented support plates therewithin, each plate having an interior opening for accommodating axial movement therethrough of the spider assembly and respective cluster of rods. The opening defining plural radially extending channels and corresponding parallel interior wall surfaces of the support plate

  6. Results of the investigations of transient fuel rod behaviour

    International Nuclear Information System (INIS)

    Fiege, A.

    1980-01-01

    The aim of the research on the fuel rod behaviour mainly effected in the KFZ Karlsruhe and at the KWU Erlangen as a part of the German reactor safety research program is to investigate the physical and chemical phenomena which are significant when the zircaloy claddings are failing, and to establish mathematical models verified by experiments by means of which the extent of damage in the reactor core in different incidents can be worked out in a realistic way. These mathematical models (program system SSYST) shall replace the conservative assumptions so far used for incident analyses and quantify their safety reserves, respectively. (orig./HP) [de

  7. The multi-interlock and check of logical system for 5 MW low power reactor automatic rod

    International Nuclear Information System (INIS)

    Li Guangjian; Zhao Zengqiao

    1992-01-01

    The safety and reliability of the logical system for 5 MW LPR automatic rod are improved, because of using multi-interlock and manual check on line. The design character and function of the logical system are introduced

  8. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 1 discusses the following topics: the background of the project; test program description; summary of tests and test results; problem evaluation; functional requirements confirmation; recommendations; and completed test documentation for tests performed in Phase 3

  9. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 9 discusses the following topics: Integrated System Normal Operations Test Results and Analysis Report; Integrated System Off-Normal Operations Test Results and Analysis Report; and Integrated System Maintenance Operations Test Results and Analysis Report

  10. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 8 discusses Control System SOT Tests Results and Analysis Report. This is a continuation of Book 7

  11. Prototypical Rod Construction Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 3 discusses the following topics: Downender Test Results and Analysis Report; NFBC Canister Upender Test Results and Analysis Report; Fuel Assembly Handling Fixture Test Results and Analysis Report; and Fuel Canister Upender Test Results and Analysis Report

  12. Mechanical components design for PWR - control rod drive mechanism

    International Nuclear Information System (INIS)

    Leme, Francisco Louzano; Mattar Neto, Miguel

    2002-01-01

    The Control Rod Drive Mechanism (CRDM) is usually - a high precision - equipment incorporating mechanical and electrical components designed to move the control rods. The 'control rods' refer to all rods or assemblies that are moved to assess the performance of the reactor. The CRDM here presented is the Nut and Lead Screw type. This type is basically a power screw type magnetically coupled to a slow speed reluctance electric motor that provides a means of axially positioning the movable fuel assemblies in the reactor core for purpose of controlling core reactivity. A helically threaded lead screw assembly, comprising one element of power screw, is attached to a movable fuel assemblies. The CRDM usually has closer and more consistent contact with environment peculiar to the reactor than has only other machinery component. This environment includes not only the radiation field of the reactor, but also the temperature, pressure and chemical properties associated with the material used as the coolant for reactor fuel. Specific and special materials are needed because of the above mentioned application. Due to the importance of the above described CRDM functions, this paper will also consider the nuclear functions and their safety classes as well as the CRDM nuclear design criteria. (author)

  13. Monitoring device for withdrawing control rods

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi.

    1985-01-01

    Purpose: To improve the sensitivity and the responsivity to an equivalent extent to those in the case where local power range monitors are densely arranged near each of the control rods, with no actual but pseudo increase of the number of local power range monitors. Constitution: The monitor arrangement is patterned by utilizing the symmetricity of the reactor core and stored in a monitor designating device. The symmetricity of control rods to be selected and withdrawn by an operator is judged by a control rod symmetry monitoring device, while the symmetricity of the withdrawn control rods is judged by a control rod withdrawal state monitoring device. Then, only when both of the devices judge the symmetricity, the control rods are subjected to gang driving by the control rod drive mechanisms. In this way, monitoring at a high sensitivity and responsivity is enabled with no increase for the number of monitors. (Yoshino, Y.)

  14. Rope wind-up type control rod

    International Nuclear Information System (INIS)

    Tsuji, Teruaki; Watanabe, Shigeru.

    1979-01-01

    Purpose: To hold a control rod at a certain position even if the sealed cover of the rod drive mechanism should fail. Constitution: A plurality of friction plates, engaging wheels and a threaded shaft are provided to the wind-up drum for winding up a rope which moves the control rod up and down. While the control rod is adapted to drop by its own weight upon insertion, it is adapted to stop at a predetermined position exactly with no shocks by gradually increasing braking force by the sliding friction caused from the friction plates or the like. A ratch mechanism is provided to the upper portion of the control rod so that the top of the ratch piece may automatically engage the guide passage wall of the control rod upon uncontrolled running of the control rod to prevent further uncontrolled running thereof. (Ikeda, J.)

  15. Hollow rods for the oil producing industry

    Energy Technology Data Exchange (ETDEWEB)

    Khalimova, L M; Elyasheva, M A

    1970-01-01

    Hollow sucker rods have several advantages over conventional ones. The hollow rods actuate the well pump and at the same time conduct produced fluids to surface. When paraffin deposition occurs, it can be minimized by injecting steam, hot oil or hot water into the hollow rod. Other chemicals, such as demulsifiers, scale inhibitors, corrosion inhibitors, etc., can also be placed in the well through the hollow rods. This reduces cost of preventive treatments, reduces number of workovers, increases oil production, and reduces cost of oil. Because the internal area of the rod is small, the passing liquids have a high velocity and thereby carry sand and dirt out of the well. This reduces pump wear between the piston and the plunger. Specifications of hollow rods, their operating characteristics, and results obtained with such rods under various circumstances are described.

  16. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 4 discusses the following topics: Rod Compaction/Loading System Test Results and Analysis Report; Waste Collection System Test Results and Analysis Report; Waste Container Transfer Fixture Test Results and Analysis Report; Staging and Cutting Table Test Results and Analysis Report; and Upper Cutting System Test Results and Analysis Report

  17. Hydraulic system for driving control rods

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1982-01-01

    Purpose: To enable safety reactor shut down upon occurrence of an abnormal excess pressure in a hydraulic control unit. Constitution: The actuation pressure for a pressure switch that generates a scram signal is set lower than the release pressure set to a pressure release valve. Thus, if the pressure of nitrogen gas in a nitrogen container increases such as upon exposure of the hydraulic control unit to a high temperature, the pressure switch is actuated at first to generate the scram signal and a scram valve is opened to supply water at high pressure to control rod drives under the driving force of the nitrogen gas at high pressure to rapidly insert the control element into the reactor and shut down it. If the pressure of the nitrogen gas still increases after the scram, the pressure release valve is opened to release the nitrogen gas at high temperature to the atmosphere. Since the scram is attained before the actuation of the pressure release valve, safety reactor shut down can be attained and the hydraulic control unit can be protected. (Sekiya, K.)

  18. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

  19. Description and characterization of the ACRR's programmable transient rod withdrawal mode

    International Nuclear Information System (INIS)

    Boldt, K.R.; Sullivan, W.H.; Kefauver, H.L.

    1980-01-01

    To satisfy experiment needs for Sandia's Advanced Reactor Safety Program, a programmable Transient Rod Withdrawal (TRW) mode has been developed for the Annular Core Research Reactor (ACRR). The programmable mode is a modification of the existing continuous-withdrawal TRW mode and permits speed and direction changes during the pulse sequence. Basically, a TRW operation is similar to a routine pulse operation except that transient rods are mechanically withdrawn rather than pneumatically fired. Being a pulse-type operation, the TRW mode complies with pulse-mode safety system settings. Control system interlocks prevent the pneumatic firing of rods in the TRW mode. The hardware for the programmable TRW mode includes three ACRR transient rods, the ACRR timer, two rod programmers, a minicomputer and a summing circuit for position indication. Each ACRR transient rod is mechanically driven by a stepping motor (rated torque at 4.24 joules) and is capable of a maximum TRW speed of 26.7 centimeters/ second. The maximum reactivity insertion rate is $2.45/second with a transient rod bank worth of $3.00 and $3.47/second with a bank worth of $4.25, which is expected to be installed soon. The ACRR timer is a multifunctional timer used in all operating modes of the reactor. In the programmable TRW mode, the timer starts the rod programmers and drops regulating rods to terminate the operation. Programmed withdrawal capability is provided by one of two rod programmers (a hardwire-based unit and a microprocessor-based unit). The hardwire unit has eight intervals in which speed, direction and distance are selected by switches on the front panel. The microprocessor-based unit has the capability of 64 intervals in which speed, direction, and distance or time can be specified. Programming this unit is accomplished from the front panel or by inputting data from an HP-9845. minicomputer via a digital I/O interface. Self-test programs in the software provide a continual check of an operating

  20. Control rod effects with plutonium recycle in a PWR

    International Nuclear Information System (INIS)

    Nash, G.; Muehl, G.J.; Gibson, I.H.

    1979-03-01

    A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

  1. Automated nuclear fuel rod pattern loading system

    International Nuclear Information System (INIS)

    Lambert, D.V.; Nylund, T.W.; Byers, J.W.; Haley, D.E. Jr.; Cioffi, J.V.

    1991-01-01

    This patent describes a method for loading fuel rods in a desired pattern. It comprises providing a supply of fuel rods of known enrichments; providing a magazine defining a matrix of elongated slots open at their forward ends for receiving fuel rods; defining a fuel rod feed path; receiving successively one at a time along the feed path fuel rods selected from the supply thereof; verifying successively one at a time along the feed path the identity of the selected fuel rods, the verifying including blocking passage of each selected fuel rod along the feed path until the identity of each selected fuel rod is confirmed as correct; feeding to the magazine successively one at a time along the feed path the selective and verified fuel rods; and supporting and moving the magazine along X-Y axes to successively align one at a time selected ones of the slots with the feed path for loading in the magazine the successive fuel rods in a desired enrichment pattern

  2. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Schmugar, K.L.

    1977-01-01

    Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

  3. SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

    2015-11-15

    Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

  4. Control rod for a reactor

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1975-01-01

    Object: To change arrangement and density of each layer of neutron absorber in the control rod and to render rotation by each layer possible, whereby the neutron absorber may be rotated to readily flatten power distribution. Structure: Neutron absorbers such as boron and carbide are filled into stainless steel pipes, which are peripherally arranged in a multi-layer fashion. Arrangement and density of the neutron absorber by each layer are changed and rotation by each layer is made possible, whereby surface area of the absorber or the like is changed to flatten power distribution. (Furukawa, Y.)

  5. Accident-tolerant control rod

    International Nuclear Information System (INIS)

    Ohta, Hirokazu; Sawabe, Takashi; Ogata, Takanari

    2013-01-01

    Boron carbide (B 4 C) and hafnium (Hf) metal are used for the neutron absorber materials of control rods in BWRs, and silver-indium-cadmium (Ag-In-Cd) alloy is used in PWRs. These materials are clad with stainless steel. The eutectic point of B 4 C and iron (Fe) is about 1150 deg. C and the melting point of Ag-In-Cd alloy is about 800 deg. C, which are lower than the temperature of zircaloy - steam reaction increases rapidly (∼1200 deg. C). Accordingly, it is possible that the control rods melt and collapse before the reactor core is significantly damaged in the case of severe accidents. Since the neutron absorber would be separated from the fuels, there is a risk of re-criticality, when pure water or seawater is injected for emergency cooling. In order to ensure sub-criticality and extend options of emergency cooling in the course of severe accidents, a concept of accident-tolerant control rod (ACT) has been derived. ACT utilises a new absorber material having the following properties: - higher neutron absorption than current control rod; - higher melting or eutectic temperature than 1200 deg. C where rapid zircaloy oxidation occurs; - high miscibility with molten fuel materials. The candidate of a new absorber material for ATC includes gadolinia (Gd 2 O 3 ), samaria (Sm 2 O 3 ), europia (Eu 2 O 3 ), dysprosia (Dy 2 O 3 ), hafnia (HfO 2 ). The melting point of these materials and the liquefaction temperature with Fe are higher than the rapid zircaloy oxidation temperature. ACT will not collapse before the core melt-down. After the core melt-down, the absorber material will be mixed with molten fuel material. The current absorber materials, such as B 4 C, Hf and Ag-In-Cd, are charged at the tip of ATC in which the neutron flux is high, and a new absorber material is charged in the low-flux region. This design could minimise the degradation of a new absorber material by the neutron absorption and the influence of ATC deployment on reactor control procedure. As a

  6. Development of a control rod drive

    International Nuclear Information System (INIS)

    1991-01-01

    In the period under review, the computer codes required for transients calculation have been completed, as well as the programs for modelling and testing the hot-gas temperature control by means of combined core rod and reflector rod operation. The specification of requirements to be fulfilled by the rod drive computer and the neutron flux measuring system has been done relying essentially on the data obtained by the transients calculations performed and the resulting informations on operating conditions. The work for optimization of the core rod drive with regard to rod driving speeds and the 'three-point switch' with hysteresis for controlled, automatic core rod operation has been concentrating on the case of specified, normal operation of the reactor. (orig./DG) [de

  7. Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Shang Zhi, E-mail: shangzhi@tsinghua.org.c [Faculty of Engineering, Kingston University, London SW15 3DW (United Kingdom); Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.co [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2010-04-15

    The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90 deg. the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  8. Expandable device for a nuclear fuel rod

    International Nuclear Information System (INIS)

    Gesinski, L.T.

    1978-01-01

    A nuclear fuel rod and a device for use within the rod cladding to maintain the axial position of the fuel pellets stacked one atop another within the cladding are described. The device is initially of a smaller external cross-section than the fuel rod cladding internal cross-section so as to accommodate loading into the rod at preselected locations. During power operation the device responds to a rise in temperature, so as to permanently maintain its position and restrain any axial motion of the fuel pellets

  9. Control rod selecting and driving device

    International Nuclear Information System (INIS)

    Isobe, Hideo.

    1981-01-01

    Purpose: To simultaneously drive a predetermined number of control rods in a predetermined mode by the control of addresses for predetermined number of control rods and read or write of driving codified data to and from the memory by way of a memory controller. Constitution: The system comprises a control rod information selection device for selecting predetermined control rods from a plurality of control rods disposed in a reactor and outputting information for driving them in a predetermined mode, a control rod information output device for codifying the information outputted from the above device and outputting the addresses to the predetermined control rods and driving mode coded data, and a driving device for driving said predetermined control rods in a predetermined mode in accordance with the codified data outputted from the above device, said control rod infromation output device comprising a memory device capable of storing a predetermined number of the codified data and a memory control device for storing the predetermined number of data into the above memory device at a predetermined timing while successively outputting the thus stored predetermined number of data at a predetermined timing. (Seki, T.)

  10. Nuclear fuel rod end plug weld inspection

    International Nuclear Information System (INIS)

    Parker, M. A.; Patrick, S. S.; Rice, G. F.

    1985-01-01

    Apparatus and method for testing TIG (tungsten inert gas) welds of end plugs on a sealed nuclear reactor fuel rod. An X-ray fluorescent spectrograph testing unit detects tungsten inclusion weld defects in the top end plug's seal weld. Separate ultrasonic weld inspection system testing units test the top end plug's seal and girth welds and test the bottom end plug's girth weld for penetration, porosity and wall thinning defects. The nuclear fuel rod is automatically moved into and out from each testing unit and is automatically transported between the testing units by rod handling devices. A controller supervises the operation of the testing units and the rod handling devices

  11. Absorber rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Acher, H.

    1985-01-01

    The invention concerns a further addition to the invention of DE 33 42 830 A1. The free contact of the hollow piston with the nut due to hydraulic pressure is replaced by a hydraulic or spring attachment. The pressure system required to produce the hydraulic pressure is therefore omitted, and the electrical power required for driving the pump or the mass flow is also omitted. The absorber rod slotted along its longitudinal axis is replaced by an absorber rod, in the longitudinal axis of which a hollow piston is connected together with the absorber rod. This makes the absorber rod more stable, and assembly is simplified. (orig./HP) [de

  12. Acoustic loading effects on oscillating rod bundles

    International Nuclear Information System (INIS)

    Lin, W.H.

    1980-01-01

    An analytical study of the interaction between an infinite acoustic medium and a cluster of circular rods is described. The acoustic field due to oscillating rods and the acoustic loading on the rods are first solved in a closed form. The acoustic loading is then used as a forcing function for rod responses, and the acousto-elastic couplings are solved simultaneously. Numerical examples are presented for several cases to illustrate the effects of various system parameters on the acoustic reaction force coefficients. The effect of the acoustic loading on the coupled eigenfrequencies are discussed

  13. Growth and Morphology of Rod Eutectics

    Energy Technology Data Exchange (ETDEWEB)

    Jing Teng; Shan Liu; R. Trivedi

    2008-03-17

    The formation of rod eutectic microstructure is investigated systematically in a succinonitrile-camphor alloy of eutectic composition by using the directional solidification technique. A new rod eutectic configuration is observed in which the rods form with elliptical cylindrical shape. Two different orientations of the ellipse are observed that differ by a 90{sup o} rotation such that the major and the minor axes are interchanged. Critical experiments in thin samples, where a single layer of rods forms, show that the spacing and orientation of the elliptic rods are governed by the growth rate and the sample thickness. In thicker samples, multi layers of rods form with circular cross-section and the scaling law between the spacing and velocity predicted by the Jackson and Hunt model is validated. A theoretical model is developed for a two-dimensional array of elliptical rods that are arranged in a hexagonal or a square array, and the results are shown to be consistent with the experimental observations. The model of elliptic rods is also shown to reduce to that for the circular rod eutectic when the lengths of the two axes are equal, and to the lamellar eutectic model when one of the axes is much larger than the other one.

  14. Nuclear reactor fuel rod attachment system

    International Nuclear Information System (INIS)

    Christiansen, D.W.

    1983-01-01

    The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

  15. Arrangement of permanent magnet and reed switches for control rod position indicator of SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. I.; Kim, J. H.; Hur, H.; Jang, M. H.

    2001-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indication. In this study, the characteristics of permanent magnet and reed switches are introduced and the calculation method for arrangement of permanent magnet and reed switch is presented

  16. Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

    2000-01-01

    Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

  17. Design of magnetic flux concentrator of permancent magnet for control rod position indicator of SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, J. Y.; Kim, J. H.; Hur, H.; Kim, J. I.

    2002-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indication. In this study, the magnetic flux concentrator of permanent magnet is introduced and the calculation method for effective flux area for reed switch is presented

  18. Dry rod consolidation technology development

    International Nuclear Information System (INIS)

    Rasmussen, T.L.; Schoonen, D.H.; Fisher, M.W.

    1986-01-01

    The Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is funding a Program to consolidate commercial spent fuel for testing in dry storage casks and to develop technology that will be fed into other OCRWM Programs, e.g., Prototypical Consolidation Demonstration Program. The Program is being conducted at the Idaho National Engineering Laboratory (INEL) by the Operating Contractor, EGandG Idaho, Inc. Hardware and software have been designed and fabricated for installation in a hot cell adjacent to the Test Area North (TAN) Hot Shop Facility. This equipment will be used to perform dry consolidation of commercial spent fuel from the Virginia Power (VP) Cooperative Agreement Spent Fuel Storage Cask (SPSC) Demonstration Program and assemblies that had previously been stored at the Engine Maintenance and Disassembly (EMAD) facility in Nevada. Consolidation will be accomplished by individual, horizontal rod pulling. A computerized semi-automatic control system with operator involvement will be utilized to conduct consolidation operations. Special features have been incorporated in the design to allow crud collection and measurement of rod pulling forces. During consolidation operations, data will be taken to characterize this technology. Still photo, video tape, and other documentation will be generated to make developed information available to interested parties. Cold checkout of the hardware and software will complete in September of 1986. Following installation in the hot cell, consolidation operations will begin in January 1987. Resulting consolidated fuel will be utilized in the VP Cooperative Agreement SFSC Program

  19. Attracting electromagnet for control rod

    International Nuclear Information System (INIS)

    Kato, Kazuo; Sasaki, Kotaro.

    1989-01-01

    Non-magnetic material plates with inherent resistivity of greater than 20 μΩ-cm and thickness of less than 3 mm are used for the end plates of attracting electromagnets for closed type control rods. By using such control rod attracting electromagnets, the scram releasing time can be shortened than usual. Since the armature attracting side of the electromagnet has to be sealed by a non-magnetic plate, a bronze plate of about 5 mm thickness has been used so far. Accordingly, non-magnetic plate is inserted to the electromagnet attracting face to increase air source length for improving to shorten the scram releasing time. This method, however, worsens the attracting property on one hand to require a great magnetomotive force. For overcoming these drawbacks, in the present invention, the material for tightly closing end plates in an electromagnet is changed from bronze plate to non-magnetic stainless steel SUS 303 or non-magnetic Monel metal and, in addition, the plate thickness is reduced to less than 5 mm thereby maintaining the attracting property and shortening the scram releasing time. (K.M.)

  20. Prototypical Rod Consolidation Demonstration Project

    International Nuclear Information System (INIS)

    1993-05-01

    The objective of Phase 3 of the Prototypical Rod consolidation Demonstration Project (PRCDP) was to procure, fabricate, assemble, and test the Prototypical Rod consolidation System as described in the NUS Phase 2 Final Design Report. This effort required providing the materials, components, and fabricated parts which makes up all of the system equipment. In addition, it included the assembly, installation, and setup of this equipment at the Cold Test Facility. During the Phase 3 effort the system was tested on a component, subsystem, and system level. This volume 1, discusses the PRCDP Phase 3 Test Program that was conducted by the HALLIBURTON NUS Environmental Corporation under contract AC07-86ID12651 with the United States Department of Energy. This document, Volume 1, Book 5 discusses the following topics: Lower Cutting System Test Results and Analysis Report; NFBC Loading System Test Results and Analysis Report; Robotic Bridge Transporter Test Results and Analysis Report; RM-10A Remotec Manipulator Test Results and Analysis Report; and Manipulator Transporter Test Results and Analysis Report

  1. Seismic scrammability of HTTR control rods

    International Nuclear Information System (INIS)

    Nishiguchi, I.; Iyoku, T.; Ito, N.; Watanabe, Y.; Araki, T.; Katagiri, S.

    1990-01-01

    Scrammability tests on HTTR (High-Temperature Engineering Test Reactor) control rods under seismic conditions have been carried out and seismic conditions influences on scram time as well as functional integrity were examined. A control rod drive located in a stand-pipe at the top of a reactor vessel, raises and lowers a pair of control rods by suspension cables. Each flexible control rod consists of 10 neutron absorber sections held together by a metal spine passing through the center. It falls into a hole in graphite blocks due to gravity at scram. In the tests, a full scale control rod drive and a pair of control rods were employed with a column of graphite blocks in which holes for rods were formed. Blocks misalignment and contact with the hole surface during earthquakes were considered as major causes of disturbance in scram time. Therefore, the following parameters were set up in the tests: excitation direction, combination or horizontal and vertical excitation, acceleration, frequency and block to block gaps. Main results obtained from tests are as follow. 1) Every scram time obtained under the design conditions was within 6 seconds. On the contrary, the scram times were 5.2 seconds when there were no vibration. Therefore, it was concluded that the seismic effects on scram time were not significant. 2) Scram time became longer with increase in both acceleration and horizontal excitation frequency, and control rods fell very smoothly without any jerkiness. This suggests that collision between control rods and hole surface is the main disturbing factor of falling motion. 3) Mechanical and functional integrity of control rod drive mechanism, control rods and graphite blocks was confirmed after 140 seismic scrammability tests. (author). 10 figs, 1 tab

  2. Hydraulic Rod Drives for the CAREM Reactor

    International Nuclear Information System (INIS)

    Mazzi, R.O

    2000-01-01

    CAREM belongs to those considered innovative reactors and their main design goal is obtain a significant improvement in safety.Requirements for the design of the first shutdown systems (FSS) is one of the mayor challenges from functional and reliability point of view, among most of the system of a nuclear reactor.Thus, the design of First Shutdown System must be in accordance with both, the system and the specific design criteria of the CAREM concept.In order to choose the best option for the control rod drive device, three different alternatives have been analysed in the frame of the Project.This paper discusses the advantages and disadvantages of each option and presents the main reasons to select the hydraulic type as the most promising one.The principles and main characteristics of the selected system are explained and the main goals to be obtained during development activities, in order to obtain a reliable design to successfully comply with operating requirements for reactor service are also presented

  3. Self-contact for rods on cylinders

    NARCIS (Netherlands)

    Heijden, van der G.H.M.; Peletier, M.A.; Planqué, R.

    2006-01-01

    We study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar variable, and the

  4. Self-contact for rods on cylinders

    NARCIS (Netherlands)

    G.H.M. van der Heijden; M.A. Peletier (Mark); R. Planqué (Robert)

    2004-01-01

    textabstractWe study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar

  5. Self-contact for rods on cylinders

    NARCIS (Netherlands)

    Heijden, van der G.H.M.; Peletier, M.A.; Planqué, R.

    2004-01-01

    We study self-contact phenomena in elastic rods that are constrained to lie on a cylinder. By choosing a particular set of variables to describe the rod centerline the variational setting is made particularly simple: the strain energy is a second-order functional of a single scalar variable, and the

  6. Tipping Time of a Quantum Rod

    Science.gov (United States)

    Parrikar, Onkar

    2010-01-01

    The behaviour of a quantum rod, pivoted at its lower end on an impenetrable floor and restricted to moving in the vertical plane under the gravitational potential, is studied analytically under the approximation that the rod is initially localized to a "small-enough" neighbourhood around the point of classical unstable equilibrium. It is shown…

  7. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  8. Spider and burnable poison rod combinations

    International Nuclear Information System (INIS)

    Walton, L.A.

    1980-01-01

    A description is given of an improved design of burnable poison rods and their associated spiders used in the fuel assemblies of pressurized water power reactor cores which allows the rods to be installed and removed more quickly, simply and gently than in previously described systems. (U.K.)

  9. Method of inspecting control rod drive mechanism

    International Nuclear Information System (INIS)

    Sato, Tomomi; Tatemichi, Shin-ichiro; Hasegawa, Hidenobu.

    1988-01-01

    Purpose: To conduct inspection for control rod drives and fuel handling operations in parallel without taking out the entire fuel, while maintaining the reactor in a subcritical state. Method: Control rod drives are inspected through the release of connection between control rods and control rod drives, detachment and dismantling of control rod drives, etc. In this case, structural materials having neutron absorbing power equal to or greater than the control rods are inserted into the gap after taking out fuels. Since the structural materials have neutron absorbing portion, subcriticality is maintained by the neutron absorbing effect. Accordingly, there is no requirement for taking out all of the fuels, thereby enabling to check the control rod drives and conduct handling for the fuels in parallel. As a result, the number of days required for the inspection can be shortened and it is possible to improve the working efficiency for the decomposition, inspection, etc. of the control rod drives and, thus, improve the operation efficiency of the nuclear power plant thereby attaining the predetermined purpose. (Kawakami, Y.)

  10. Control rod guide tube assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

    An improved fuel assembly is described as consisting of a sleeve that engages one end of a control rod guide tube essentially fixing the guide tube to one of the fuel assembly end structures. The end of the sleeve protrudes above the surface of the end fitting. The outer surface of the sleeve has a peripheral groove that engages the resilient sides of a cellular grid or lattice shaped lock. This lock fixes the sleeve in position between the various elements that comprise the end fitting, thereby eliminating a profusion of costly and potentially troublesome nuts, threaded studs and the like that are frequently employed in the fuel assemblies that are presently in use

  11. Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de.

    1984-01-01

    In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

  12. Development of joining techniques for fabrication of fuel rod simulators

    International Nuclear Information System (INIS)

    Moorhead, A.J.; McCulloch, R.W.; Reed, R.W.; Woodhouse, J.J.

    1980-10-01

    Much of the safety-related thermal-hydraulic tests on nuclear reactors are conducted not in the reactor itself, but in mockup segments of a core that uses resistance-heated fuel rod simulators (FRS) in place of the radioactive fuel rods. Laser welding and furnace brazing techniques are described for joining subassemblies for FRS that have survived up to 1000 h steady-state operation at 700 to 1100 0 C cladding temperatures and over 5000 thermal transients, ranging from 10 to 100 0 C/s. A pulsed-laser welding procedure that includes use of small-diameter filler wire is used to join one end of a resistance heating element of Pt-8 W, Fe-22 Cr-5.5 Al-0.5 Co, or 80 Ni-20 Cr (wt %) to a tubular conductor of an appropriate intermediate material. The other end of the heating element is laser welded to an end plug, which in turn is welded to a central conductor rod

  13. Nuclear reactor with scrammable part length rod

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

  14. RODMOD: a code for control rod positioning

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1978-11-01

    The report documents a computer code which has been implemented to position control rods according to a prescribed schedule during the calculation of a reactor history. Control rods may be represented explicitly with or without internal black absorber conditions in selected energy groups, or fractional insertion may be done, or both, in a problem. There is provision for control rod follower, movement of materials through a series of zones in a closed loop, and shutdown rod insertion and subsequent removal to allow the reactor history calculation to be continued. This code is incorporated in the system containing the VENTURE diffusion theory neutronics and the BURNER exposure codes for routine use. The implemented automated procedures cause the prescribed control rod insertion schedule to be applied without the access of additional user input data during the calculation of a reactor operating history

  15. Vortex Noise from Rotating Cylindrical Rods

    Science.gov (United States)

    Stowell, E Z; Deming, A F

    1935-01-01

    A series of round rods of the some diameter were rotated individually about the mid-point of each rod. Vortices are shed from the rods when in motion, giving rise to the emission of sound. With the rotating system placed in the open air, the distribution of sound in space, the acoustical power output, and the spectral distribution have been studied. The frequency of emission of vortices from any point on the rod is given by the formula von Karman. From the spectrum estimates are made of the distribution of acoustical power along the rod, the amount of air concerned in sound production, the "equivalent size" of the vortices, and the acoustical energy content for each vortex.

  16. Microcomputer system for controlling fuel rod length

    International Nuclear Information System (INIS)

    Meyer, E.R.; Bouldin, D.W.; Bolfing, B.J.

    1979-01-01

    A system is being developed at the Oak Ridge National Laboratory (ORNL) to automatically measure and control the length of fuel rods for use in a high temperature gas-cooled reactor (HTGR). The system utilizes an LSI-11 microcomputer for monitoring fuel rod length and for adjusting the primary factor affecting length. Preliminary results indicate that the automated system can maintain fuel rod length within the specified limits of 1.940 +- 0.040 in. This system provides quality control documentation and eliminates the dependence of the current fuel rod molding process on manual length control. In addition, the microcomputer system is compatible with planned efforts to extend control to fuel rod fissile and fertile material contents

  17. Heat transfer in rod bundles with severe clad deformations

    International Nuclear Information System (INIS)

    Ihle, P.

    1984-04-01

    The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL) [de

  18. Stepping movement analysis of control rod drive mechanism

    International Nuclear Information System (INIS)

    Xu Yantao; Zu Hongbiao

    2013-01-01

    Background: Control rod drive mechanism (CRDM) is one of the important safety-related equipment for nuclear power plants. Purpose: The operating parameters of stepping movement, including lifting loads, step distance and step velocity, are all critical design targets. Methods: FEA and numerical simulation are used to analyze stepping movement separately. Results: The motion equations of the movable magnet in stepping movement are established by load analysis. Gravitation, magnetic force, fluid resistance and spring force are all in consideration in the load analysis. The operating parameters of stepping movement are given. Conclusions: The results, including time history curves of force, speed and etc, can positively used in the design of CRDM. (authors)

  19. Express diagnostics of WWER fuel rods at nuclear power plants

    International Nuclear Information System (INIS)

    Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

    2009-01-01

    Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

  20. Measurement of control rods efficiency at the RB reactor by pulse method; Merenje efikasnosti kontrolnih sipki u reaktoru RB impulsnom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Markovic, V; Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1963-07-01

    Pulse method was applied for measuring the efficiency of control rods at the RB reactor. This paper describes the theory of experiment, experimental procedure applied and results obtained. Results are considered to be useful for safety analysis. it was found that the influence of delayed neutrons is rather small and could be neglected in estimation of rods efficiency.

  1. Fabrication Of Control Rod System Of The RSG-GAS

    International Nuclear Information System (INIS)

    Sudirdjo, Hari; Setyono; Prasetya, Hendra

    2001-01-01

    Eight units of control rod mechanical system of RSG-GAS has been fabricated. The control rod mechanical system of RSG-GAS consist of guide tube and lifting rod. Complete construction of the control rod mechanical system of RSG-GAS are guide tube, lifting rod, absorber, and absorber casing. The eight units of the control rod mechanical system of RSG-GAS has been fabricated according to the mechanical engineering design

  2. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  3. BWR feedwater nozzle and control-rod-drive return line nozzle cracking

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems

  4. A user input manual for single fuel rod behaviour analysis code FEMAXI-III

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

    1983-03-01

    Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

  5. Failure Mode and Effects Analysis (FMEA) of the solid state full length rod control system

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Full Length Rod Control System (FLRCS) controls the power to the rod drive mechanisms for rod movement in response to signals received from the Reactor Control System or from signals generated through Reactor Operator action. Rod movement is used to control reactivity of the reactor during plant operation. The Full Length Rod Control System is designed to perform its reactivity control function in conjunction with the Reactor Control and Protection System, to maintain the reactor core within design safety limits. By the use of a Failure Mode and Effects Analysis, it is shown that the FLRCS will perform its reactivity control functions considering the loss of single active components. That is, sufficient fault limiting control circuits are provided which blocks control rod movement and/or indicates presence of a fault condition at the Control Board. Reactor operator action or automatic reactor trip will thus mitigate the consequences of potential failure of the FLRCS. The analysis also qualitatively demonstrates the reliability of the FLRCS to perform its intended function

  6. Nondestructive assay of HTGR fuel rods

    International Nuclear Information System (INIS)

    Menlove, H.O.

    1974-01-01

    Performance characteristics of three different radioactive source NDA systems are compared for the assay of HTGR fuel rods and stacks of rods. These systems include the fast neutron Sb-Be assay system, the 252 Cf ''Shuffler,'' and the thermal neutron PAPAS assay system. Studies have been made to determinethe perturbation on the measurements from particle size, kernel Th/U ratio, thorium content, and hydrogen content. In addition to the total 235 U determination, the pellet-to-pellet or rod-to-rod uniformity of HTGR fuel rod stacks has been measured by counting the delayed gamma rays with a NaI through-hole in the PAPAS system. These measurements showed that rod substitutions can be detected easily in a fuel stack, and that detailed information is available on the loading variations in a uniform stack. Using a 1.0 mg 252 Cf source, assay rates of 2 to 4 rods/s are possible, thus facilitating measurement of 100 percent of a plant's throughput. (U.S.)

  7. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  8. French LMFBR's control rods experience and development

    International Nuclear Information System (INIS)

    Arnaud, G.; Guigon, A.; Verset, L.

    1983-06-01

    Since the last ten years, the French program has been, first of all, directed to the setting up, and then the development of, at once, the Phenix control rods, and next, the Super-Phenix ones. The vented pin design, with porous plug and sodium bonding, which allows the choices of large diameters, has been taken, since the Rapsodie experience was decisive. The absorber material is sintered, 10 B enriched, boron carbide. The can is made of 316 type stainless steel, stabilised, or not, with titanium. The experience gained in Phenix up to now is important, and deals with about six loads of control rods. Results confirm the validity of the design of the absorber pins. Some difficulties has been encountered for the guiding devices, due to the swelling of the steel. They have required design and material improvements. Such difficulties are discarded by a new design of the bearing, for the Super-Phenix control rods. The other parts of these rods, from the Primary Shut-Down System, are strictly derived from Phenix. The design of the rods from the Secondary Shut-Down System is rather different, but it's not the case for the design of the absorber pins: in many a way, they are derived from Phenix pins and from Rapsodie control rods. Both types of rods irradiation tests are in progress in Phenix [fr

  9. High-yield production of hydrophobins RodA and RodB from Aspergillus fumigatus in Pichia pastoris

    DEFF Research Database (Denmark)

    Pedersen, Mona Højgaard; Borodina, Irina; Moresco, Jacob Lange

    2011-01-01

    A as well as rRodB were able to convert a glass surface from hydrophilic to hydrophobic similar to native RodA, but only rRodB was able to decrease the hydrophobicity of a Teflon-like surface to the same extent as native RodA, while rRodA showed this ability to a lesser extent. Recombinant RodA and native...

  10. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  11. Control rod housing alignment and repair apparatus

    International Nuclear Information System (INIS)

    Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

    1991-01-01

    This patent describes a welding a repair device for precisely locating and welding the position of the top of a control rod drive housing attached from a stub tube from a corresponding aperture and alignment pin in a core plate within a boiling water nuclear reactor, the welding and repair device. It comprises: a shaft, the shaft extending from the vicinity of the top of the control rod drive housing up to and through the aperture in the core plate; means for registering to the aperture and the alignment pin on the core plate; a fixture attached to the bottom end of the shaft for mating to the top of the control rod drive housing in precise mating relationship; the fixture attached to the bottom end of the shaft whereby the fixture, when mated to the control rod drove housing and the registering means when registered to the alignment pin and aperture on the core plate imparts to the shaft, and angularity between the top of the control rod drive housing and the hole in the core plate; a hollow cylinder, the cylinder mounted for depending and sealed support with respect to the shaft above, about and below the control rod drive housing top; the cylinder depending down below the control rod drive housing to an elevation below the top of the sub tube; a rotating welding apparatus with a welding head for dispensing weldment mounted for rotation with respect to the shaft; the welding head disposed at the juncture between the side of the control rod drive housing and the stub tube; and means for flooding the cylinder with gas whereby the cylinder may be lowered. flooded in a gas environment and effect a weld between the top of the stub tube and the control rod drive housing

  12. Control rod housing alignment and repair method

    International Nuclear Information System (INIS)

    Dixon, R.C.; Deaver, G.A.; Punches, J.R.; Singleton, G.E.; Erbes, J.G.; Offer, H.P.

    1992-01-01

    This patent describes a method for underwater welding of a control rod drive housing inserted through a stub tube to maintain requisite alignment and elevation of the top of the control rod drive housing to an overlying and corresponding aperture in a core plate as measured by an alignment device which determines the relative elevation and angularity with respect to the aperture. It comprises providing a welding cylinder dependent from the alignment device such that the elevation of the top of the welding cylinder is in a fixed relationship to the alignment device and is gas-proof; pressurizing the welding cylinder with inert welding gas sufficient to maintain the interior of the welding cylinder dry; lowering the welding cylinder through the aperture in the core plate by depending the cylinder with respect to the alignment device, the lowering including lowering through and adjusting the elevation relationship of the welding cylinder to the alignment device such that when the alignment device is in position to measure the elevation and angularity of the new control rod drive housing, the lower distal end of the welding cylinder extends below the upper periphery of the stub where welding is to occur; inserting a new control rod drive housing through the stub tube and positioning the control rod drive housing to a predetermined relationship to the anticipated final position of the control rod drive housing; providing welding implements transversely rotatably mounted interior of the welding cylinder relative to the alignment device such that the welding implements may be accurately positioned for dispensing weldment around the periphery of the top of the stub tube and at the side of the control rod drive housing; measuring the elevation and angularity of the control rod drive housing; and dispensing weldment along the top of the stub tube and at the side of the control rod drive housing

  13. The effect of aging upon CE and B and W control rod drives

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.

    1991-01-01

    Though mechanically different, the control rod drive (CRD) systems used at both Combustion Engineering (CE) and Babcock and Wilcox (B and W) plants position the control rod assemblies (CRA) in the core in response to automatic or manual reactivity control signals. Both systems are also designed to provide a rapid insertion of the CRAs upon a loss of AC power. The CRD system consists of the actual drive mechanisms, power and control, rod position indication, and cooling system components. This aging evaluation included the individual absorber rods, and the fuel assembly and upper internal guide tubes, since failure of these components could preclude the insertion of the control assemblies. Aging and environmental degradation have resulted in system and component failures. Many of these failures caused dropped or slipped rods which adversely affected plant operations by resulting in power reductions, scrams, and safety system actuation. No CRD system failure has ever resulted in the inability to shut down a reactor. However, unplanned, automatic trips challenge the operation of the plants safety systems. Consequently, their occurrence represents a potentially significant increase in plant risk. System and component failures have resulted in four Information Notices during the past decade

  14. Apparatus for handling control rod drives

    International Nuclear Information System (INIS)

    Akimoto, A.; Watanabe, M.; Yoshida, T.; Sugaya, Z.; Saito, T.; Ishii, Y.

    1979-01-01

    An apparatus for handling control rod drives (CRD's) attached by detachable fixing means to housings mounted in a reactor pressure vessel and each coupled to one of control rods inserted in the reactor pressure vessel is described. The apparatus for handling the CRD's comprise cylindrical housing means, uncoupling means mounted in the housing means for uncoupling each of the control rods from the respective CRD, means mounted on the housing means for effecting attaching and detaching of the fixing means, means for supporting the housing means, and means for moving the support means longitudinally of the CRD

  15. Depletion calculations of adjuster rods in Darlington

    Energy Technology Data Exchange (ETDEWEB)

    Arsenault, B.; Tsang, K., E-mail: benoit.arsenault@amecfw.com, E-mail: kwok.tsang@amecfw.com [AMEC Foster Wheeler, Toronto, ON (Canada)

    2015-07-01

    This paper describes the simulation methodology and reactivity worth calculated for aged adjuster rods in the Darlington core. ORIGEN-S IST was applied to simulate the isotope transmutation process of the stainless steel and titanium adjusters. The compositions were used in DRAGON-IST to calculate the change in incremental properties of aged adjusters. Pre-simulations of the reactivity worth of the stainless steel and titanium adjusters in Darlington were performed using RFSP-IST and the results showed that the titanium adjuster rods exhibit faster reactivity-worth drop than that of stainless steel rods. (author)

  16. LOFT fuel rod pressure measurement

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-01-01

    Pressure sensors selected for measuring fuel rod pressure within the LOFT reactor exhibited stable, repeatable operating characteristics during calibrations at temperatures up to 800 0 F and pressures to 2500 psig. All sensors have a nominal sensitivity of .5 millivolts per psi, decreasing monotonically with temperature. Output signal increases linearly with increasing pressure up to 2000 psig. For imposed slow and rapid temperature variations and for pressure applied during these tests, the sensor indicates a pressure at variance with the actual value by up to 15% of reading. However, the imposed temperature rates of change often exceeded the value of -10 0 F/sec. specified for LOFT. The series of tests in an autoclave permit creation of an environment most closely resembling sensor operating conditions within LOFT. For multiple blowdowns and for longtime durations the sensor continued to provide pressure-related output signals. For temperature rates up to -87 0 F/sec, the indicated pressure measurement error remained less than 13% of reading. Adverse effects caused by heating the 1/16 inch O.D. signal cable to 800 0 F contributed only insignificantly to the noted pressure measurement error

  17. Dry rod consolidation technology development

    International Nuclear Information System (INIS)

    Rasmussen, T.L.; Schoonen, D.H.; Feldman, E.M.; Fisher, M.W.

    1987-01-01

    The Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is funding a program to consolidate commercial spent fuel for testing in dry storage casks and to develop technology that will be fed into other OCRWM programs, e.g., Prototypical Consolidation Demonstration Program (PCDP). The program is being conducted at the Idaho National Engineering Laboratory (INEL) by the INEL Operating Contractor EG and G Idaho, Inc. Hardware and software have been designed and fabricated for installation in a hot cell adjacent to the Test Area North (TAN) Hot Shop Facility. This equipment is used to perform dry consolidation of commercial spent fuel from the Virginia Power (VP) Cooperative Agreement Spent Fuel Storage Cask (SFSC) Demonstration Program and assemblies that had previously been stored at the Engine Maintenance and Disassembly (EMAD) facility in Nevada. Consolidation is accomplished by individual, horizontal rod pulling. A computerized semiautomatic control system with operator involvement is utilized to conduct consolidation operations. During consolidation operations, data is taken to characterize this technology. Still photo, video tape, and other documentation will be generated to make developed information available to interested parties. Cold checkout of the hardware and software was completed in September of 1986. Following installation in the hot cell, consolidation operations begins in May 1987. Resulting consolidated fuel will be utilized in the VP Cooperative Agreement SFSC Program

  18. Water rod and fuel assembly

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Tada, Nobuo; Nakajima, Junjiro; Aizawa, Yasuhiro.

    1995-01-01

    A water rod disposed in a fuel assembly comprises a larger diameter tube constituting an upwarding flow channel for coolants flown from the lower portion of a reactor core, and a smaller diameter tube connected fixedly to the larger diameter tube at the periphery of the upper end thereof and constituting a downwarding flow channel for coolants upwardly flown in the larger diameter tube. The larger diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to α + β treatment for once. The smaller diameter tube is formed by subjecting a base tube made of a zirconium alloy to PILGER mil fabrication and annealing in α region repeatingly for several times, then subjecting it to β treatment for once. With such procedures, the amount of irradiation growth of the tube in the axial direction is made greater in the larger diameter tube than that in the smaller diameter tube. Accordingly, since the smaller diameter tube is never bent by pressing, mechanical integrity of the fuel assembly is never lost. (I.N.)

  19. Taylor impact of glass rods

    International Nuclear Information System (INIS)

    Willmott, G.R.; Radford, D.D.

    2005-01-01

    The deformation and fracture behavior of soda-lime and borosilicate glass rods was examined during classic and symmetric Taylor impact experiments for impact pressures to 4 and 10 GPa, respectively. High-speed photography and piezoresistive gauges were used to measure the failure front velocities in both glasses, and for impact pressures below ∼2 GPa the failure front velocity increases rapidly with increasing pressure. As the pressure was increased above ∼3 GPa, the failure front velocities asymptotically approached maximum values between the longitudinal and shear wave velocities of each material; at ∼4 GPa, the average failure front velocities were 4.7±0.5 and 4.6±0.5 mm μs -1 for the soda-lime and borosilicate specimens, respectively. The observed mechanism of failure in these experiments involved continuous pressure-dependent nucleation and growth of microcracks behind the incident wave. As the impact pressure was increased, there was a decrease in the time to failure. The density of cracks within the failed region was material dependent, with the more open-structured borosilicate glass showing a larger fracture density

  20. Control rod drive hydraulic device

    International Nuclear Information System (INIS)

    Takekawa, Toru.

    1994-01-01

    The device of the present invention can reliably prevent a possible erroneous withdrawal of control rod driving mechanism when the pressure of a coolant line is increased by isolation operation of hydraulic control units upon periodical inspection for a BWR type reactor. That is, a coolant line is connected to the downstream of a hydraulic supply device. The coolant line is connected to a hydraulic control unit. A coolant hydraulic detection device and a pressure setting device are disposed to the coolant line. A closing signal line and a returning signal line are disposed, which connect the hydraulic supply device and a flow rate control valve for the hydraulic setting device. In the device of the present invention, even if pressure of supplied coolants is elevated due to isolation of hydraulic control units, the elevation of the hydraulic pressure can be prevented. Accordingly, reliability upon periodical reactor inspection can be improved. Further, the facility is simplified and the installation to an existent facility is easy. (I.S.)

  1. High-resolution flow structure measurements in a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Ylönen, A. T.

    2013-07-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  2. High-resolution flow structure measurements in a rod bundle

    International Nuclear Information System (INIS)

    Ylönen, A. T.

    2013-01-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  3. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  4. Evolution of fuel rod support under irradiation consequences on the mechanical behavior of fuel assembly

    International Nuclear Information System (INIS)

    Billerey, A.; Bouffioux, P.

    2002-01-01

    The complete paper follows. According to the fuel management policy in French PWR with respect to high burn-up, the prediction of the mechanical behavior of the irradiated fuel assembly is required as far as excessive deformations of fuel assembly might lead to incomplete Rod Cluster Control Assembly insertion (safety problems) and fretting wear lead to leaking rods (plant operation problems). One of the most important parameter is the evolution of the fuel rod support in the grid cell as it directly governs the mechanical behavior of the fuel assembly and consequently allows to predict the behavior of irradiated structure in terms of (i) axial and lateral deformation (global behavior of the assembly) and (ii) fretting wear (local behavior of the rod). Fuel rod support is provided by a spring-dimple system fixed on the grid. During irradiation, the spring force decreases and a gap between the rod and the spring might open. This phenomenon is due to (i) irradiation-induced stress relaxation for the spring and for the dimples, (ii) grid growth and (iii) reduction of rod diameter. Two models have been developed to predict the behavior of the rod in the grid cell. The first model is able to evaluate the spring force relaxation during irradiation. The second one is able to evaluate the rotation characteristic of the fuel rod in the cell, function of the spring force. The main input parameters are (i) the creep laws of the grid materials, (ii) the growth law of the grid, (iii) the evolution of rod diameter and (iv) the design of the fuel rod support. The objectives of this paper are to: (i) evaluate the consequences of grid support design modifications on the fretting sensitivity in terms of predicted maximum gap during irradiation and operational time to gap appearance; (ii) evaluate, using a non-linear Finite Element assembly model, the impact of the evolution of grid support under irradiation on the mechanical behavior of the full assembly in terms of axial and

  5. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1976-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilent members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  6. Control rod position detector for nuclear reactor

    International Nuclear Information System (INIS)

    Kudo, Mitsuru; Fujiwara, Hiroshi.

    1981-01-01

    Purpose: To improve the reliability of a control rod position detector by detecting a reactive code with a combination of control rod position change signals produced from vertical and horizontal axis decoders, generation an error signal and thus simultaneously detecting the operation of more than two lead switches. Constitution: Horizontal and vertical axis position signals responsive to changes in the control rod position are applied from lead switches connected in a predetermined matrix connection corresponding to the notches of the positions of respective position detecting probes, the reactive output from the decoder is detected by a reactive code detecting circuit, which in turn generates a fault signal, and the control rod position code converted in a notch number generating circuit is converted to a predetermined value indicating invalidity. Accordingly, a fault caused by the simultaneous operation of a plurality of failed lead switches can be effectively detected. (Yoshino, Y.)

  7. Control rod for a nuclear reactor

    Science.gov (United States)

    Roman, Walter G.; Sutton, Jr., Harry G.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

  8. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

    Yoder, G.L.; Morris, D.G.; Mullins, C.B.

    1985-01-01

    Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

  9. Genetics Home Reference: cone-rod dystrophy

    Science.gov (United States)

    ... common cause of autosomal recessive cone-rod dystrophy , accounting for 30 to 60 percent of cases. At ... dystrophy play essential roles in the structure and function of specialized light receptor cells (photoreceptors) in the ...

  10. Stabilizing device for control rod tip

    International Nuclear Information System (INIS)

    Verdone, G.F.

    1982-01-01

    A control rod has a spring device on its lower end for eliminating oscillatory contact of the rod against its adjacent guide tube wall. The base of the device is connected to the lower tip of the rod. A plurality of elongated extensions are cantilevered downward from the base. Each extension has a shoulder for contacting the guide tube, and the plurality of shoulders as a group has a transverse dimension that is preset to be larger than the inner diameter of the guide tube such that an interference fit is obtained when the control rod is inserted in the tube. The elongated extensions form an open-ended, substantially hollow member through which most of the liquid coolant flows, and the spaces between adjacent extensions allow the flow to bypass the shoulders without experiencing a significant pressure drop

  11. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  12. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  13. Intrasacral rod fixation for pediatric lumbopelvic fusion.

    Science.gov (United States)

    Ilharreborde, Brice; Mazda, Keyvan

    2014-07-01

    This paper reports the authors' 19 years experience with pediatric intrasacral rod fixation. After insertion of two cannulated screws in S1 with and an original template guiding them into the anterior third of the endplate, two short fusion rods were inserted into the sacrum according to Jackson's technique distally to S3. In neuromuscular scoliosis, pelvic obliquity was reduced by connecting the proximal and distal constructs, distraction or compression, and in situ rod bending. In children with high-grade spondylolisthesis, lumbosacral kyphosis was reduced by rotation of the sacrum and in situ bending. There were no direct neurological or vascular injuries. The main complication was infection (7%). No pseudarthrosis or significant loss of correction at the lumbosacral junction was observed during follow-up. Intrasacral rod fixation appears to be safe and reliable for lumbopelvic fusion in pediatric patients.

  14. Cutting system for burnable poison rod

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Toyama, Norihide; Koshino, Yasuo; Fujii, Toshio

    1989-01-01

    Burnable poison rods attached to spent fuels are contained in a containing box and transported to a receiving pool. The burnable poison rod-containing box is provisionally situated by the operation to a handling device to a provisional setting rack in a cutting pool and attached to a cutting guide of a cutting device upon cutting. The burnable poison rod is cut only in a cutting pool water and tritium generated upon cutting is dissolved into the cutting pool water. Diffusion of tritium is thus restricted. Further, the cutting pool is isolated by a partition device from the receiving pool during cutting of the burnable poison rod. Accordingly, water in which tritium is dissolved is inhibited from moving to the receiving pool and prevail of tritium contamination can be avoided. (T.M.)

  15. Detection device for control rod scram

    International Nuclear Information System (INIS)

    Ishiyama, Satoshi.

    1989-01-01

    The device of the present invention comprises a control rod dropping separately from a control rod driving mechanism main body, a following tube falling separately accompanying therewith and a guide tube for guiding the dropping of the control rod and the following tube. Further, rare earth permanent magnets are embedded with the pole being axially oriented in the following tube and bobbins each mounted with an inner flange made of high magnetic permeability material are disposed to the guide tube. Coils are wound in the bobbin. In this control rod scram detection device, since magnetic fluxes can effectively be supplied to the coils, it is possible to obtain stable and highly reliable scram detection signals. Further, since the coils and the bobbins can be manufactured separately from the guide tube, their assemblies can be tested independently from the guide tube. (K.M.)

  16. Safety logic systems of PFBR

    International Nuclear Information System (INIS)

    Sambasivan, S. Ilango

    2004-01-01

    Full text : PFBR is provided with two independent, fast acting and diverse shutdown systems to detect any abnormalities and to initiate safety action. Each system consists of sensors, signal processing systems, logics, drive mechanisms and absorber rods. The absorber rods of the first system are Control and Safety Rods (CSR) and that of the second are called as Diverse Safety Rods (DSR). There are nine CSR and three DSR. While CSR are used for startup, control of reactor power, controlled shutdown and SCRAM, the DSR are used only for SCRAM. The respective drive mechanisms are called as CSRDM and DSRDM. Each of these two systems is capable of executing the shutdown satisfactorily with single failure criteria. Two independent safety logic systems based on diverse principles have been designed for the two shut down systems. The analog outputs of the sensors of Core Monitoring Systems comprising of reactor flux monitoring, core temperature monitoring, failed fuel detection and core flow monitoring systems are processed and converted into binary signals depending on their instantaneous values. Safety logic systems receive the binary signals from these core-monitoring systems and process them logically to protect the reactor against postulated initiating events. Neutronic and power to flow (P/Q) signals form the inputs to safety logic system-I and temperature signals are inputs to the safety logic system II. Failed fuel detection signals are processed by both the shut down systems. The two logic systems to actuate the safety rods are also based on two diverse designs and implemented with solid-state devices to meet all the requirements of safety systems. Safety logic system I that caters to neutronic and P/Q signals is designed around combinational logic and has an on-line test facility to detect struck at faults. The second logic system is based on dynamic logic and hence is inherently safe. This paper gives an overview of the two logic systems that have been

  17. Freely suspended rod fall dampener, especially for control rod of liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Becvar, J.; Saroch, V.

    1977-01-01

    A shock absorber is described whose advantage is that the space required for the movement of the shock absorber in the operating travel of the system suspension rod-control rod bundle may be reduced. The design allows the automatic disconnection of the system and the removal of the suspension rod from the reactor without dismantling. The braking force reaction is transmitted to the structure above the core. The system fall energy is absorbed on the side of the suspension rod which has a bigger mass. (J.B.)

  18. Conceptual design of stepper motor replacing servo motor for control rod controller

    International Nuclear Information System (INIS)

    Mohd Dzul Aiman Aslan; Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat

    2010-01-01

    In PUSPATI TRIGA Reactor, current control rod controller are using servo motor to control the movement. Control rod is a very important safety element and measure in every nuclear reactor. So, precision is very important in measurement of security in the nuclear reactor. In this case, there are a few disadvantages when using the servo motor is measurement of the motor is not precise. One solution to overcome this is by shifting servo motor with stepper motor. A stepper motor (or step motor) is a brush less, synchronous electric motor that can divide a full rotation into a large number of steps. (author)

  19. Development of the predictive maintenance system prototype for the rod control system

    International Nuclear Information System (INIS)

    Lim, H. S.; Hong, H. P.; Koo, J. M.; Kim, Y. B.; Han, H. W.

    2003-01-01

    The demand for safety and reliability of Nuclear Power Plants (NPPs) has been constantly increasing and economical operation is also an important issue. Developing and adopting predictive maintenance technology for the major systems or equipment is considered as a way to achieve these goals. This paper describes the development of a predictive maintenance system prototype for the Rod Control System, which adopts an advanced methodology. Bayesian Belief Networks (BBN) has been adopted for the real time fault diagnosis and prediction of the system. Through a simulation test, it was confirmed that the prototype monitors and secures sound operability of rod drive mechanism and its control system, and also provides the predictive maintenance information

  20. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

    In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

  1. Control rod supporting device in reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Itooka, Satoshi; Harada, Kiyoshi; Jodoi, Takashi.

    1990-01-01

    Since coolants flowing from a reactor core hit against a control rod and a control rod connection pipe, a considerable amount of bending moment for separating an attracting surface between an electromagnet and an armature is formed. Then, a plurality of grooves are formed on a heat sensitive material to dispose a heat collecting fin, and each of upper and lower contact portions of a control rod supporting portion in which the flanged portion of T-like cross section does not slip out is made into a partial spheric surface and a portion between the electromagnet and the attracted member are engaged by the unevenness. With such a constitution, even if a bending moment is applied, the control rod only swings and the bending moment is not transmitted to the attracted member. Further, since the temperature of the heat sensitive material can be rapidly made closer to the peripheral temperature by using the heat collecting fin, the timing for separation is made accurate. Further, since the engaging portion is brought into contact at the spheric surface, the load distribution on the control rod is made uniform, and the positional relationship is made accurate, to support the control rod reliably and the separation depends only on the temperature of the coolants. (N.H.)

  2. Control rod excess withdrawal prevention device

    International Nuclear Information System (INIS)

    Takayama, Yoshihito.

    1992-01-01

    Excess withdrawal of a control rod of a BWR type reactor is prevented. That is, the device comprises (1) a speed detector for detecting the driving speed of a control rod, (2) a judging circuit for outputting an abnormal signal if the driving speed is greater than a predetermined level and (3) a direction control valve compulsory closing circuit for controlling the driving direction of inserting and withdrawing a control rod based on an abnormal signal. With such a constitution, when the with drawing speed of a control rod is greater than a predetermined level, it is detected by the speed detector and the judging circuit. Then, all of the direction control valve are closed by way of the direction control valve compulsory closing circuit. As a result, the operation of the control rod is stopped compulsorily and the withdrawing speed of the control rod can be lowered to a speed corresponding to that upon gravitational withdrawal. Accordingly, excess withdrawal can be prevented. (I.S)

  3. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

  4. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  5. Control device for a nuclear reactor with a multitude of control rods, extending into the reactor core from above, with linear drive mechanisms and additional gripper devices

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    The components of the additional gripper devices with magnetically operated finger-shaped latches are separated from the also magnetically operated latches of the linear drive mechanisms in order to avoid common-mode failures when fast shutdown is required. Only part of the safety rods are held by the additional gripping devices in the withdrawn position. There is provided for recording elements indicating positively which one of the safety locks is gearing with the control rods. At the upper end of each control rod there is a coupling head held by electromagnetically operated locking devices in the withdrawn position, if control power is available. (DG) [de

  6. Effect of local automatic control rods on three-dimensional calculations of the power distribution in an RBMK

    International Nuclear Information System (INIS)

    Pogosbekyan, L.R.; Lysov, D.A.; Bronitskii, L.L.

    1993-01-01

    Numerical simulators and information systems that support nuclear reactor operators must have fast models to estimate how fuel reloads and control rod displacement affect neutron and power distributions in the core. The consequences of reloads and control rod displacement cannot be evaluated correctly without considering local automatic control-rod operations in maintaining the radial power distribution. Fast three-dimensional models to estimate the effects of reloads and displacement of the control and safety rods have already been examined. I.V. Zonov et al. used the following assumptions in their calculational model: (1) the full-scale problem could be reduced a three-dimensional fragment of a locally perturbed core, and (2) the boundary conditions of the fragment and its total power were constant. The last assumption considers approximately how local automatic control rods stabilize the radial power distribution, but three dimensional calculations with these rods are not considered. These assumptions were introduced to obtain high computational speed. I.L. Bronitskii et al. considered in more detail how moving the local automatic control rods affect the power dimensional in the three-dimensional fragment, because, with on-line monitoring of the reload process, information on control rod positions is periodically renewed, and the calculations are done in real time. This model to predict the three-dimensional power distribution to (1) do a preliminary reload analysis, and (2) prepare the core for reloading did not consider the effect of perturbations from the local automatic control rods. Here we examine a model of a stationary neutron distribution. On one hand it gives results in an acceptable computation time; on the other it is a full-scale three-dimensional model and considers how local automatic control rods affect both the radial and axial power distribution

  7. Safety, Efficacy and Acceptability of Implanon a Single Rod ...

    African Journals Online (AJOL)

    However at 12 months there were significant but within normal reductions (p < 0.003/0.05) in the systolic and diastolic blood pressures. The side effects were menstrual abnormalities. Eighteen (56.3%), 1 (3.1%) and 13 (40.6%) reported reduced, increased and combinations of bleeding patterns respectively. No participant ...

  8. Sensitivity studies for 3-D rod ejection analyses on axial power shape

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min-Ho; Park, Jin-Woo; Park, Guen-Tae; Ryu, Seok-Hee; Um, Kil-Sup; Lee, Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-10-15

    The current safety analysis methodology using the point kinetics model combined with numerous conservative assumptions result in unrealistic prediction of the transient behavior wasting huge margin for safety analyses while the safety regulation criteria for the reactivity initiated accident are going strict. To deal with this, KNF is developing a 3-D rod ejection analysis methodology using the multi-dimensional code coupling system CHASER. The CHASER system couples three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST using message passing interface (MPI). A sensitivity study for 3-D rod ejection analysis on axial power shape (APS) is carried out to survey the tendency of safety parameters by power distributions and to build up a realistic safety analysis methodology while maintaining conservatism. The currently developing 3-D rod ejection analysis methodology using the multi-dimensional core transient analysis code system, CHASER was shown to reasonably reflect the conservative assumptions by tuning up kinetic parameters.

  9. Nondestructive examination of irradiated fuel rods by pulsed eddy current techniques

    International Nuclear Information System (INIS)

    Francis, W.C.; Quapp, W.J.; Martin, M.R.; Gibson, G.W.

    1976-02-01

    A number of fuel rods and unfueled zircaloy cladding tubes which had been irradiated in the Saxton reactor have undergone extensive nondestructive and corroborative destructive examinations by Aerojet Nuclear Company as part of the Water Reactor Safety Research Program, Irradiation Effects Test Series. This report discusses the pulsed eddy current (PEC) nondestructive examinations on the fuel rods and tubing and the metallography results on two fuel rods and one irradiated zircaloy tube. The PEC equipment, designed jointly by Argonne National Laboratory and Aerojet, performed very satisfactorily the functions of diameter, profile, and wall thickness measurements and OD and ID surface defect detection. The destructive examination provided reasonably good confirmation of ''defects'' detected in the nondestructive examination

  10. The thermo-mechanics of the PWR fuel rod; La thermomecanique du crayon de combustible REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France, EDF, 75 - Paris (France); Gautier, B.; Chaigne, G. [Electricite de France, Service Etudes et Projets Thermiques et Nucleaires, 75 - Paris (France)] [and others

    1999-03-29

    The fuel rod mechanics is of a great importance in the safety and performance of the reactors. In this domain a meeting has been organized by the SFEN the 18 march 1998 at Paris. With the participation of scientists from CEA, EDF and Framatome, the physics of the fuel rods was presented based on four main aspects. Two first papers dealt with the solicitations of the fuel rod in normal and accidental conditions. The physical phenomena under irradiation were then detailed in the four following talks. Three papers presented the simulation and the codes of the fuel-cladding interactions with the diabolo effect. The last paper was devoted to the experiment feedback and the research programs. (A.L.B.)

  11. Elliptical cross section fuel rod study II; Estudio de barras combustibles de seccion eliptica II

    Energy Technology Data Exchange (ETDEWEB)

    Taboada, H; Marajofsky, A [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Combustibles Nucleares

    1997-12-31

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab.

  12. The development of the fuel rod transient performance analysis code FTPAC

    International Nuclear Information System (INIS)

    Han Zhijie; Ji Songtao

    2014-01-01

    Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

  13. Mechanical reliability of falling systems. Case of control rods of nuclear reactors

    International Nuclear Information System (INIS)

    Rezkallah, B.

    1991-05-01

    After having briefly recalled the main principles adopted for nuclear reactor safety (barrier based method, emergency stop system, emergency cooling system) and how control rod operation anomalies are described, this research thesis first gives an overview of knowledge in the field of reliability (principles, qualitative and quantitative methods, failure trees, performance function, shock modelling and mode analysis). Then, the author presents the modelling of the drop and rotation of a control rod. The computation technique is the modal synthesis. The author highlights the influence of physical (friction, initial conditions) and numerical (modal truncation) parameters. He reports a probabilistic analysis of the problem and proposes a stochastic model for the different shock forces (notably friction). He compares the results with those obtained with the DEMT software. The last part reports the investigation of a rare case: the jamming or untimely stop of the control rod device

  14. Design characteristics of metallic fuel rod on its in-LMR performance

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

  15. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  16. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  17. Temperature distribution on fuel rods: a study on the effect of eccentricity in the position of UO2 pellets

    International Nuclear Information System (INIS)

    Gaspar Junior, Joao Carlos Aguiar

    2010-01-01

    This work proposes the development of a method of solving equations of heat transfer applied in fuel rods using the finite element method, in order to evaluate the performance and safety of the nuclear system. Was prepared in a Fortran program to evaluate the equations governing the problem, the boundary conditions and apply the properties of materials on a steady state. This program uses the mesh generation input and graphical output generated by the program GID. The method was validated against the analytical solution found in the book Todreas and Kazimi with error less than 0.2% and with respect to the improved analytical solution of Nijsing for axisymmetry rod and eccentricity rod with error less than a 3.6%. Applications have been developed with the use of correlations for properties with the temperature dependence of resolution axisymmetry rod and the resolution of a rod with eccentricity. The method developed, should it be implemented, would allow the assessment of fuel rods in the given situations and other scenarios, as well as adding a tool of substantial value in the analysis of rods. (author)

  18. Design, construction and simulation of a multipurpose system for precision movement of control rods in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shirazi, S.A. Mousavi, E-mail: a_moosavi@azad.ac.i [Department of Physics, Islamic Azad University, South Tehran Branch, Tehran (Iran, Islamic Republic of); Aghanajafi, C. [Department of Mechanical Engineering, K.N.T. University of Technology, Tehran (Iran, Islamic Republic of); Sadoughi, S. [Department of Electrical Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Sharifloo, N. [Department of Physics, Imam Hossein University, Tehran (Iran, Islamic Republic of)

    2010-12-15

    This article presents the design and implementation of a microcontroller-based system for the automatic movement of control rods in nuclear reactors of either power or research types. This system is controlled automatically, is linked to a personal computer system, and has manual controlling ability as well. The important features of this system are: automatic scram of the control rods, activation of alarm in emergency situations, and the ability to tune the control rod movement course both upwards and downwards. In this system, a small tank has been improvised as a coolant reservoir for pool type reactors such as Tehran Research Reactor and its water level is continuously adjusted by special sensors. Also, this system can be applied for controlling various types of control rods such as the regulating rods, safety rods and shim rods; can be connected to all reactor measurement tools and systems such as the period meter, power meter and flux meter; and can receive feedback signals from them. The devised system can be calibrated with these measurement tools by two special potentiometers in the related electronic board. The processes of this system have been simulated by the SIMULINK tool kit of MATLAB software and all responses of the system, including oscillation and transient responses, have been analyzed.

  19. RODDRP - A FORTRAN program for use in control rod calibration by the rod drop method

    International Nuclear Information System (INIS)

    Wilson, W.E.

    1972-01-01

    The different methods to measure reactivity which are applicable to control rod calibration are discussed. They include: 1) the positive period method, 2) the rod drop method, 3) the source-jerk method, 4) the rod oscillation method, and 5) the pulsed neutron method. The instrument setup used at WSU for rod drop measurements is presented. To speed up the analysis of power fall-off trace, a FORTRAN IV program called RODDRP was written to simultaneously solve the in-hour equation and relative neutron flux. The procedure for calculating the worth of the rod that produced the power trace is given. The reactivity for each time relative flux point is obtained. Conclusions about the status of the equipment are made

  20. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  1. Pressure loss in two-phase flow through a microchannel rod bundle

    International Nuclear Information System (INIS)

    Smith, A.C.; Hamm, L.L.; Qureshi, Z.; Steeper, T.J.

    1998-01-01

    The purpose of the microchannel rod bundle two-phase flow test described here was to provide data for benchmarking safety analyses for the accelerator production of tritium (APT). The objective was to obtain pressure loss data for a typical accelerator target rod bundle over a wide range of two-phase flow conditions. The test rod bundle assembly was fabricated for single-phase pressure drop tests conducted at Los Alamos National Laboratory (LANL) and subsequently used for the two-phase flow testing described here. The results for a typical case are given. These results fall generally in the slug flow regime for the horizontal flow results of Fukano and Kariyasaki for a 1.0-mm circular channel. Fukano and Kariyasaki found that surface tension effects were dominant in the 1-mm channel and report no churn regime. The results were also compared with the flow regime maps given by Triplett et al. for flow in discrete microchannels. Triplett employed both circular and trapezoidal channels, the latter to approximate the rod bundle interstitial flow channel shape. It was found that the rod bundle flow fell across the slug-to-churn flow regime transition reported by Triplett. This is consistent with the expectation that cross flow among channels would result in turbulent mixing and would suppress the formation of large discrete bubbles

  2. Performance estimation of control rod position indicator due to aging of magnet

    International Nuclear Information System (INIS)

    Yu, Je Yong; Kim, Ji Ho; Huh, Hyung; Choi, Myoung Hwan; Sohn, Dong Seong

    2009-01-01

    The Control Element Drive Mechanism (CEDM) for the integral reactor is designed to raise and lower the control rod in steps of 2mm in order to satisfy the design features of the integral reactor which are the soluble boron free operation and the use of a nuclear heating for the reactor start-up. The actual position of the control rod could be achieved to sense the magnet connected to the control rod by the position indicator around the upper pressure housing of CEDM. It is sufficient that the actual position information of control rod at 20mm interval from the position indicator is used for the core safety analysis. As the magnet moves upward along the position indicator assembly from the bottom to the top in the upper pressure housing, the output voltage increases linearly step-wise at 0.2VDC increments. Between every step there are transient areas which occur by a contact closing of three reed switches which is the 2-3-2 contact closing sequence. In this paper the output voltage signal corresponding to the position of control rod was estimated on the 2-1-2 contact closing sequence due to the aging of the magnet.

  3. Control rod drop accident analysis for the mixed core project in Ling Ao NPS

    International Nuclear Information System (INIS)

    Zhang Shishun; Zhou Zhou; Xiao Min

    2004-01-01

    AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

  4. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  5. Preliminary study for the transport of the fuel rods of U235 enriched to 1.8 per cent

    International Nuclear Information System (INIS)

    Cardenas, H.; Perez, A.

    1998-01-01

    Transport of 1,8% U235 enriched fuel rods needs both the evaluation of the radiological risk and considerations about criticality aspects. Issues as diverse production characteristics, storage facilities in the source of origin an economical aspects have to be added to the radiological and nuclear considerations. Transport of those rods through national territory must comply with the Argentine Regulatory authority's regulations, based on the Safety Series No. 6, (ed. 1985) -as amended 1990- IAEA. Safety criteria are exposed, taking into account the amount of material to be transported, container characteristics, packaging type and expedition conditions. (author)

  6. Gelation And Mechanical Response of Patchy Rods

    Science.gov (United States)

    Kazem, Navid; Majidi, Carmel; Maloney, Craig

    We perform Brownian Dynamics simulations to study the gelation of suspensions of attractive, rod-like particles. We show that details of the particle-particle interactions can dramatically affect the dynamics of gelation and the structure and mechanics of the networks that form. If the attraction between the rods is perfectly smooth along their length, they will collapse into compact bundles. If the attraction is sufficiently corrugated or patchy, over time, a rigid space spanning network forms. We study the structure and mechanical properties of the networks that form as a function of the fraction of the surface that is allowed to bind. Surprisingly, the structural and mechanical properties are non-monotonic in the surface coverage. At low coverage, there are not a sufficient number of cross-linking sites to form networks. At high coverage, rods bundle and form disconnected clusters. At intermediate coverage, robust networks form. The elastic modulus and yield stress are both non-monotonic in the surface coverage. The stiffest and strongest networks show an essentially homogeneous deformation under strain with rods re-orienting along the extensional axis. Weaker, clumpy networks at high surface coverage exhibit relatively little re-orienting with strong non-affine deformation. These results suggest design strategies for tailoring surface interactions between rods to yield rigid networks with optimal properties. National Science Foundation and the Air Force Office of Scientific Research.

  7. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

    The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

  8. Control rod driving hydraulic pressure device

    International Nuclear Information System (INIS)

    Ishida, Kazuo.

    1990-01-01

    Discharged water after actuating control rod drives in a BWR type reactor is once discharged to a discharging header, then returned to a master control unit and, subsequently, discharged to a reactor by way of a cooling water header. The radioactive level in the discharging header and the master control unit is increased by the reactor water to increase the operator's exposure. In view of the above, a riser is disposed for connecting a hydraulic pressure control unit incorporating a directional control valve and the cooling water head. When a certain control rod is inserted, the pressurized driving water is supplied through a hydraulic pressure control unit to the control rod drives. The discharged water from the control rod drives is entered by way of the hydraulic pressure control unit into the cooling water header and then returned to the reactor by way of other hydraulic pressure control unit and the control rod drives. Thus, the reactor water is no more recycled to the master control unit to reduce the radioactive exposure. (N.H.)

  9. Management of radioactive disused lightning rods

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Paulo de Oliveira; Silva, Fabio, E-mail: pos@cdtn.br, E-mail: silvaf@cdtn.br [Centro de Desenvolvimento da Energia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of {sup 241}Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the {sup 241}Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

  10. Gray rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Francis, T.A.; Cerni, Samuel.

    1986-01-01

    The invention relates to an improved gray rod for insertion in a nuclear fuel assembly having an array of fuel rods. The gray rod includes a thin-walled cladding tube a first longitudinal section of which is positioned within, and a second longitudinal section of which is positioned essentially without, the array of fuel rods when the gray rod is inserted in the fuel assembly. The first longitudinal section defines a pellet-receiving space having detained therein a stack of annular pellets with an outer diameter sufficient to lend radial support to the wall of the first longitudinal tube section. The second longitudinal section defines a hollow space devoid of pellets and having means to resist radial collapse under external pressure. This means may be a partially compressed spiral spring which serves the dual purpose of retaining the stack of pellets in the pellet-receiving space and of lending radial support to the wall of the second longitudinal tube section or it may be holes through the wall to allow pressure equalisation. The cladding tube is composed of stainless-steel material having a low neutron-capture cross-section, and the annular pellets preferably being composed of Zircaloy or Zirconia material. (author)

  11. Broadband Vibration Attenuation Using Hybrid Periodic Rods

    Directory of Open Access Journals (Sweden)

    S. Asiri

    2008-12-01

    Full Text Available This paper presents both theoretically and experimentally a new kind of a broadband vibration isolator. It is a table-like system formed by four parallel hybrid periodic rods connected between two plates. The rods consist of an assembly of periodic cells, each cell being composed of a short rod and piezoelectric inserts. By actively controlling the piezoelectric elements, it is shown that the periodic rods can efficiently attenuate the propagation of vibration from the upper plate to the lower one within critical frequency bands and consequently minimize the effects of transmission of undesirable vibration and sound radiation. In such a system, longitudinal waves can propagate from the vibration source in the upper plate to the lower one along the rods only within specific frequency bands called the "Pass Bands" and wave propagation is efficiently attenuated within other frequency bands called the "Stop Bands". The spectral width of these bands can be tuned according to the nature of the external excitation. The theory governing the operation of this class of vibration isolator is presented and their tunable filtering characteristics are demonstrated experimentally as functions of their design parameters. This concept can be employed in many applications to control the wave propagation and the force transmission of longitudinal vibrations both in the spectral and spatial domains in an attempt to stop/attenuate the propagation of undesirable disturbances.

  12. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  13. Radiological characterization of spent control rod assemblies

    International Nuclear Information System (INIS)

    Lepel, E.A.; Robertson, D.E.; Thomas, C.W.; Pratt, S.L.; Haggard, D.L.

    1995-10-01

    This document represents the final report of an ongoing study to provide radiological characterizations, classifications, and assessments in support of the decommissioning of nuclear power stations. This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determined by three separate methodologies, including (1) direct assay techniques, (2) calculational techniques, and (3) by sampling and laboratory radiochemical analyses. For the BWR control rod blade (CRB) and PWR burnable poison rod assembly (BPRA), 60 Co and 63 Ni, present in the stainless steel cladding, were the most abundant neutron activation products. The most abundant radionuclide in the PWR rod cluster control assembly (RCCA) was 108m Ag (130 yr halflife) produced in the Ag-In-Cd alloy used as the neutron poison. This radionuclide will be the dominant contributor to the gamma dose rate for many hundreds of years. The results of the direct assay methods agree very well (±10%) with the sampling/radiochemical measurements. The results of the calculational methods agreed fairly well with the empirical measurements for the BPRA, but often varied by a factor of 5 to 10 for the CRB and the RCCA assemblies. If concentration averaging and encapsulation, as allowed by 10CFR61.55, is performed, then each of the entire control assemblies would be classified as Class C low-level radioactive waste

  14. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  15. Overview of Japanese control rods development program

    International Nuclear Information System (INIS)

    Koyama, M.

    1984-01-01

    The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

  16. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  17. Management of radioactive disused lightning rods

    International Nuclear Information System (INIS)

    Santos, Paulo de Oliveira; Silva, Fabio

    2013-01-01

    The manufacture of radioactive lightning rod was allowed from 1970 to 1989. This authorization was based on state-of-the art science of that time that verified that radioactive lightning rods had efficiency superior to the conventional lightning rods, denominated Franklin. However, the experience showed that their efficiency was not superior enough to justify the use of radioactive sources. Consequently, in 1989, the National Commission or Nuclear Energy - CNEN, issued the Resolution 04/89 from 04-19-1989, that forbidden the importation of 241 Am tapes, assembling and commercialization of radioactive lightning-rods. The institutes of CNEN are responsible for receiving these lightning-rods and sending to the users procedures for removing and dispatch to the institutes. Therewith, these devices are kept away from the human being and environment. The Nuclear technology Development Center - CDTN and Institute for Energy and Nuclear Research - IPEN of CNEN, has built laboratories appropriate for dismantling such devices and store the 241 Am tapes safely. Nowadays are being researched methodologies to evaluate the contamination levels of the frame for possible recycling and become better the management of these devices. (author)

  18. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    Schulz, W.

    2000-01-01

    concept - to calculate the control rod's working life both in the control position and the shut-down position - will automatically lead to an optimization of the control rod strategy. Control rod optimisation is demonstrated by accumulating the total amount of control rods required in a medium-sized BWR up to the total reactor holding period. At least 60% of the first core inventory - for this control rod type an existing EMPIRICAL MODEL is already available - may be used up to the total operating period without any safety loss. Looking to the present disposal situation this concept represents a practical way to reduce all high level waste. In addition benefit of utilizing this concept is that it minimizes tritium emission. Control-rods utilized within Boiling Water Reactors (BWR) are designed for the purpose to control and shape the neutron flux profile in the reactor, to adjust the range of regulation referring to the weight rate of the reactor coolant and thirdly by- shutting down the reactor at any time and under any conditions with regard to nuclear aspects, mechanical integrity and control rod history. The designation control- or shut down rod characterize the particular field of activity for a given control rod. The focal point of my work had shown to be a calculation of the nuclear working life of any control rod design as well as an optimisation method with reference to the holding period for a given control rod inventory as a result of measuring data and a theoretical analysis describing the parameters in a general validity form. (author)

  19. Thermal-stress analysis of HTGR fuel and control rod fuel blocks in in-block carbonization and annealing furnace

    International Nuclear Information System (INIS)

    Gwaltney, R.C.; McAfee, W.J.

    1977-01-01

    The equivalent solid plate method, in conjunction with two-dimensional plane stress and plane strain analyses, was used in assessing the thermal stress behavior of HTGR fuel and control rod fuel blocks. For the control rod fuel blocks, particular attention was given to ascertaining the effects of the reserve shutdown hole and the control rod channel holes. The assumed safety factor of 2 on the failure criteria was considered adequate to account for neglecting the axial temperature gradient in the plane analyses of the ends of the blocks. The analyses indicated that the maximum calculated tensile stress values were smaller than the criteria values except for the plane strain analysis of the control rod fuel block end surfaces and the axisymmetric analysis of the fuel block as a circular cylinder. However, most of the maximum calculated strain values were greater than the criteria values

  20. 3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

    International Nuclear Information System (INIS)

    Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2016-01-01

    Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

  1. Analysis of magnetic field and hysteresis of reed switches for control rod position indicator of SMART CEDM

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, J. H.; Heo, H.; Kim, J. I.; Jang, M. H.

    2002-01-01

    The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indication. In this study, the hysteresis of reed switches is introduced and the design method using the magnetic analysis of reed switches in presented

  2. Method and apparatus for inspection of nuclear fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1977-01-01

    A method and apparatus are provided for the inspection of nuclear fuel rods to detect defects or failures in such rods. Assemblies of fuel rods are immersed in water and means are provided for causing a change in the relative pressures in the water and within the fuel rod such that fluid is expelled from the rod through any defects that may exist. Means are also provided for thereafter vibrating the rods to cause additional internal fluid or other material that may be trapped in the rod to be expelled. Sensors are provided for detecting the emission of bubbles of fluid or other material from the rod and for locating the position of the defective rod in the assembly. 5 figures

  3. Rod consolidation at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab

  4. Method and apparatus for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system requiring periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described comprising the steps of: (1) removing the top end from pulling members having electrodes of weld elements in leading ends thereof in sequence through a fuel rod container and thence through respective consolidating passages in a fuel-rod directing chamber; (3) welding the weld elements of the pulling members to the top end of respective fuel rods corresponding to the respective pulling members; (4) drawing each of the pulling members axially to draw the respective engaged fuel rods in one axial direction through the respective passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another to the one axial direction into the fuel rod container while maintaining the compacting configuration in a fuel rod density which is greater than that of the fuel rod density of the fuel rod assembly

  5. The turbulent flow in rod bundles

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1989-01-01

    Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained [pt

  6. A nuclear reactor with buffered control rods

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1974-01-01

    The control rods for, e.g., water-cooled reactors are fastened as units on common crossbars in vertical downward direction. The fastening on the crossbar is achieved by means of cross-shaped parts, e.g., in the shape of a double 'H'. A cylinder connected with a drive rod in normal operation is joined to each of the crossbars. In an emergency shut-down, this connection is interrupted and the control rod unit drops into the core through the action of gravity. Its fall is slowed down by a cushion or shock absorbing unit. For this purpose a piston is provided mounted on the supporting plate below the cylinder and guided within it. In the cylinder, the coolant is contained as damping medium. An upper opening in the cylinder serves as a ventilation hole. The movement of the piston is limited by a stopping part within the cylinder and slowed down by a spiral spring. (DG) [de

  7. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  8. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  9. Flame spread along thermally thick horizontal rods

    Science.gov (United States)

    Higuera, F. J.

    2002-06-01

    An analysis is carried out of the spread of a flame along a horizontal solid fuel rod, for which a weak aiding natural convection flow is established in the underside of the rod by the action of the axial gradient of the pressure variation that gravity generates in the warm gas surrounding the flame. The spread rate is determined in the limit of infinitely fast kinetics, taking into account the effect of radiative losses from the solid surface. The effect of a small inclination of the rod is discussed, pointing out a continuous transition between upward and downward flame spread. Flame spread along flat-bottomed solid cylinders, for which the gradient of the hydrostatically generated pressure drives the flow both along and across the direction of flame propagation, is also analysed.

  10. PBF/LOFT Lead Rod Test Program experiment operating specification

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.

    1978-11-01

    The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Understanding the behavior of light-water reactors (LWR) under loss-of-coolant conditions is a major objective of the NRC Reactor Safety Research Program. The Loss of Fluid Test (LOFT) facility is the major testing facility to evaluate the systems response of an LWR over a wide range of Loss of Coolant Experment (LOCE) conditions. As such, the LOFT core is intended to be used for sequential LOCE tests provided no significant fuel rod failures occur. The PFB/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-2 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel over a wide range of power. Thus, a relatively accurate assessment of the state of the LOFT core after the completion of each subtest and the anticipated effect of the next test can be obtained by utilizing a combination of LLR test data and analytical predictions. Specifications for the test program are presented

  11. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  12. Equipment to weld fuel rods of mixed oxides

    International Nuclear Information System (INIS)

    Aparicio, G.; Orlando, O.S.; Olano, V.R.; Toubes, B.; Munoz, C.A.

    1987-01-01

    Two welding outfits system T1G were designed and constructed to weld fuel rods with mixed oxides pellets (uranium and plutonium). One of them is connected to a glove box where the loading of sheaths takes place. The sheaths are driven to the welder through a removable plug pusher in the welding chamber. This equipment was designed to perform welding tests changing the parameters (gas composition and pressure, welding current, electrode position, etc.). The components of the welder, such as plug holder, chamber closure and peripheral accessories, were designed and constructed taking into account the working pressures in the machine, which is placed in a controlled area and connected to a glove box, where special safety conditions are necessary. The equipment to weld fuel bars is complemented by another machine, located in cold area, of the type presently used in the fuel elements factory. This equipment has been designed to perform some welding operations in sheaths and mixed oxide rods of the type Atucha I and II. Both machines have a programmed power supply of wide range and a vacuum, and pressurizing system that allows the change of parameters. Both systems have special features of handling and operation. (Author)

  13. Digital, electromagnetic rod position indicator with compensation

    International Nuclear Information System (INIS)

    Feilchenfeld, M.M.; Geis, C.G.

    1985-01-01

    A digital rod position indicator having discrete coils L 0 , L 1 , L 2 ..... spaced along the travel path of an elongate magnetically permeable member stores in digital form compensation signals for automatically adjusting the location relative to the coils at which a digital output signal representative of the position of the end of the elongate member transitions from one code to the next. The appropriate compensation signal is addressed using the digital output signal and a correction factor which takes into account the direction of movement including reversals. Reference is made to the positioning of the control rods in a pressurized water reactor. (author)

  14. Environmental report for rod storage. Volume V

    International Nuclear Information System (INIS)

    Danese, F.L.

    1981-09-01

    Volume V of this report examines the environmental impact of the rod consolidation program. The postulated, nonsite-specific, differential impacts are primarily additional occupational exposures due to the rod storage operations. Other potential radiological and nonradiological impacts that are identified and addressed are negligible. There are no increases in population exposures except those associated with transportation of spent fuel and waste material. The increased utilization of existing spent fuel storage space could result in a decrease in the nonrecoverable material resources lost to new permanent spent fuel storage

  15. Oligo(naphthylene–ethynylene) Molecular Rods

    DEFF Research Database (Denmark)

    Cramer, Jacob Roland; Ning, Yanxiao; Shen, Cai

    2013-01-01

    of palladium-catalyzed Sonogashira reactions between naphthyl halides and acetylenes. The triazene functionality was used as a protected iodine precursor to allow linear extension of the molecular rods during the synthe-ses. The carboxylic acid groups in the target molecules were protected as esters during......Molecular rods designed for surface chirality studies have been synthesized in high yields. The molecules are composed of oligo(naphthylene–ethynylene) skeletons and functionalized at their two termini with carboxylic acids and hydrophobic groups. The molecular skeletons were constructed by means...

  16. Effects of different rod spacers (helical types) on coolant crossmixing

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

    1981-11-01

    The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

  17. Processing of poison rods with a view to disposal

    International Nuclear Information System (INIS)

    Bichet, R.; Charamathieu, A.; Lasseur, C.; Golicheff, I.; Pouteaux, M.

    1979-01-01

    In the core of the French 900 and 1300 MW reactors, a certain number of rods have to be processed as wastes, particularly the burnable poison rods used during reactor start-up (900 MW: 68 rods; 1300 MW: 96 rods). Several solutions are possible: cutting and conditionning in reactor pool; transfer of the poison rods to a cutting and conditionning facility; transfer of the poison rods and fuel assemblies to a storage area where they are cutted and stored. Each of these solutions are studied, the advantages and disadvantages being presented

  18. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  19. Ejected control rod and rods drop measurements during Mochovce startup physical tests

    International Nuclear Information System (INIS)

    Minarcin, Miroslav; Elko, Marek

    1998-01-01

    Paper deals with measurements of asymmetric reactivity insertion into the reactor core that were carried out during physical startup tests of Mochovce Unit 1 in June 1998. Control rods worth measurements with one and two rods s tucked in upper limit and worth measurement of one control rod from group 6 'ejected' from the reactor core are discussed. During the experiments neutron flux was measured by four ionisation chambers (three of them were placed symmetrically around the reactor core). Results of measurements and influence of asymmetric reactivity influence on ionisation chambers response are presented in the paper. (Authors)

  20. Mechanical properties of bioresorbable self-reinforced posterior cervical rods.

    Science.gov (United States)

    Savage, Katherine; Sardar, Zeeshan M; Pohjonen, Timo; Sidhu, Gursukhman S; Eachus, Benjamin D; Vaccaro, Alexander

    2014-04-01

    A biomechanical study. To test the mechanical and physical properties of self-reinforced copolymer bioresorbable posterior cervical rods and compare their mechanical properties to commonly used Irene titanium alloy rods. Bioresorbable instrumentation is becoming increasingly common in surgical spine procedures. Compared with metallic implants, bioresorbable implants are gradually reabsorbed as the bone heals, transferring the load from the instrumentation to bone, eliminating the need for hardware removal. In addition, bioresorbable implants produce less stress shielding due to a more physiological modulus of elasticity. Three types of rods were used: (1) 5.5 mm copolymer rods and (2) 3.5 mm and (3) 5.5 mm titanium alloy rods. Four tests were used on each rod: (1) 3-point bending test, (2) 4-point bending test, (3) shear test, and (4) differential scanning calorimeter test. The outcomes were recorded: Young modulus (E), stiffness, maximum load, deflection at maximum load, load at 1.0% strain of the rod's outer surface, and maximum bending stress. The Young modulus (E) for the copolymer rods (mean range, 6.4-6.8 GPa) was significantly lower than the 3.5 mm titanium rods (106 GPa) and the 5.5 mm titanium rods (95 GPa). The stiffness of the copolymer rods (mean range, 16.6-21.4 N/mm) was also significantly lower than the 3.5 mm titanium alloy rods (43.6 N/mm) and the 5.5 mm titanium alloy rods (239.6 N/mm). The mean maximum shear load of the copolymer rods was 2735 N and they had significantly lower mean maximum loads than the titanium rods. Copolymer rods have adequate shear resistance, but less load resistance and stiffness compared with titanium rods. Their stiffness is closer to that of bone, causing less stress shielding and better gradual dynamic loading. Their use in semirigid posterior stabilization of the cervical spine may be considered.

  1. Modeling and simulation performance of sucker rod beam pump

    Energy Technology Data Exchange (ETDEWEB)

    Aditsania, Annisa, E-mail: annisaaditsania@gmail.com [Department of Computational Sciences, Institut Teknologi Bandung (Indonesia); Rahmawati, Silvy Dewi, E-mail: silvyarahmawati@gmail.com; Sukarno, Pudjo, E-mail: psukarno@gmail.com [Department of Petroleum Engineering, Institut Teknologi Bandung (Indonesia); Soewono, Edy, E-mail: esoewono@math.itb.ac.id [Department of Mathematics, Institut Teknologi Bandung (Indonesia)

    2015-09-30

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.

  2. Modeling and simulation performance of sucker rod beam pump

    International Nuclear Information System (INIS)

    Aditsania, Annisa; Rahmawati, Silvy Dewi; Sukarno, Pudjo; Soewono, Edy

    2015-01-01

    Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research

  3. RodPilotR - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

    International Nuclear Information System (INIS)

    Baron, Clemens

    2008-01-01

    With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

  4. RodPilot{sup R} - The Innovative and Cost-Effective Digital Control Rod Drive Control System for PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baron, Clemens [AREVA NP GmbH, NLEE-G, Postfach 1199, 91001 Erlangen (Germany)

    2008-07-01

    With RodPilot, AREVA NP offers an innovative and cost-effective system for controlling control rods in Pressurized Water Reactors. RodPilot controls the three operating coils of the control rod drive mechanism (lift, moveable gripper and stationary gripper coil). The rods are inserted into or withdrawn from the core as required by the Reactor Control System. The system combines modern components, state-of-the-art logic and a proven electronic control rod drive control principle to provide enhanced reliability and lower maintenance costs. (author)

  5. ROD meeting: Striving for excellence

    International Nuclear Information System (INIS)

    Taylor, G.M.

    1993-01-01

    At the 16th Reactor Operations International Topical Meeting, held August 15-18, in Melville, Long Island, N.Y., speakers pointed to the continued need to improve the operation of nuclear power plants in the United States. The meeting was sponsored by the American Nuclear Society's Reactor Operations Division and Long Island Local Section. Session speakers pointed out that plant safety is in large measure in the hands of reactor operators independent of the hardware and software in place for reactor operations. Others touched on the economic costs that nuclear energy is facing, compared with coal. Direct labor constitutes 55 percent of the power production costs of nuclear, compared to only 17 percent for coal. The industry has to recognize the opportunity to maximize labor use and reduce costs. Further improving the cost-effectiveness of energy generation will require reducing the dollar investment in plant operation. Speakers proposed reorganizational efforts for the nuclear industry, which will provide a more directed and competitive industry able to compete in todays market

  6. Patient-specific Distraction Regimen to Avoid Growth-rod Failure.

    Science.gov (United States)

    Agarwal, Aakash; Jayaswal, Arvind; Goel, Vijay K; Agarwal, Anand K

    2018-02-15

    A finite element study to establish the relationship between patient's curve flexibility (determined using curve correction under gravity) in juvenile idiopathic scoliosis and the required distraction frequency to avoid growth rod fracture, as a function of time. To perform a parametric analysis using a juvenile scoliotic spine model (single mid-thoracic curve with the apex at the eighth thoracic vertebra) and establish the relationship between curve flexibility (determined using curve correction under gravity) and the distraction interval that allows a higher factor of safety for the growth rods. Previous studies have shown that frequent distraction with smaller magnitude of distractions are less likely to result in rod failure. However there has not been any methodology or a chart provided to apply this knowledge on to the individual patients that undergo the treatment. This study aims to fill in that gap. The parametric study was performed by varying the material properties of the disc, hence altering the axial stiffness of the scoliotic spine model. The stresses on the rod were found to increase with increased axial stiffness of the spine, and this resulted in the increase of required optimal frequency to achieve a factor of safety of two for growth rods. A relationship between the percentage correction in Cobb's angle due to gravity alone, and the required distraction interval for limiting the maximum von Mises stress to 255 MPa on the growth rods was established. The distraction interval required to limit the stresses to the selected nominal value reduces with increase in stiffness of the spine. Furthermore, the appropriate distraction interval reduces for each model as the spine becomes stiffer with time (autofusion). This points to the fact the optimal distraction frequency is a time-dependent variable that must be achieved to keep the maximum von Mises stress under the specified factor of safety. The current study demonstrates the possibility of

  7. Connectedness percolation of hard deformed rods

    NARCIS (Netherlands)

    Drwenski, Tara; Dussi, Simone; Dijkstra, Marjolein; van Roij, Rene; van der Schoot, Paul

    2017-01-01

    Nanofiller particles, such as carbon nanotubes or metal wires, are used in functional polymer composites to make them conduct electricity. They are often not perfectly straight cylinders but may be tortuous or exhibit kinks. Therefore we investigate the effect of shape deformations of the rod-like

  8. Piston rod seal for a Stirling engine

    Science.gov (United States)

    Shapiro, Wilbur

    1984-01-01

    In a piston rod seal for a Stirling engine, a hydrostatic bearing and differential pressure regulating valve are utilized to provide for a low pressure differential across a rubbing seal between the hydrogen and oil so as to reduce wear on the seal.

  9. Automatic operation device for control rods

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1984-01-01

    Purpose: To enable automatic operation of control rods based on the reactor operation planning, and particularly, to decrease the operator's load upon start up and shutdown of the reactor. Constitution: Operation plannings, demand for the automatic operation, break point setting value, power and reactor core flow rate change, demand for operation interrupt, demand for restart, demand for forecasting and the like are inputted to an input device, and an overall judging device performs a long-term forecast as far as the break point by a long-term forecasting device based on the operation plannings. The automatic reactor operation or the like is carried out based on the long-term forecasting and the short time forecasting is performed by the change in the reactor core status due to the control rod operation sequence based on the control rod pattern and the operation planning. Then, it is judged if the operation for the intended control rod is possible or not based on the result of the short time forecasting. (Aizawa, K.)

  10. Confinement stabilises single crystal vaterite rods.

    OpenAIRE

    Schenk, AS; Albarracin, EJ; Kim, YY; Ihli, J; Meldrum, FC

    2014-01-01

    Single-crystals of vaterite, the least-stable anhydrous polymorph of CaCO3, are rare in biogenic and synthetic systems. We here describe the synthesis of high aspect ratio single crystal vaterite rods under additive-free conditions by precipitating CaCO3 within the cylindrical pores of track-etch membranes.

  11. CNEN resolution phohibits radioactive lightning rods

    International Nuclear Information System (INIS)

    1989-01-01

    After 15 years of irrestricted use in Brazil, the radioactive lightning rods were phohibited by Brazilian CNEN since the publication of a new law (Resolution number 4 of april 19,1989) published on may 9, 1989. All the existing ones will be removed at the time of their programed maintenance. (A.C.A.S.) [pt

  12. On contact numbers in random rod packings

    NARCIS (Netherlands)

    Wouterse, A.; Luding, Stefan; Philipse, A.P.

    2009-01-01

    Random packings of non-spherical granular particles are simulated by combining mechanical contraction and molecular dynamics, to determine contact numbers as a function of density. Particle shapes are varied from spheres to thin rods. The observed contact numbers (and packing densities) agree well

  13. Morphogenesis of rod-shaped sacculi

    NARCIS (Netherlands)

    den Blaauwen, T.; de Pedro, M.A.; Nguyen-Distèche, M.; Ayala, J.A.

    2008-01-01

    For growth and division of rod-shaped bacteria, the cylindrical part of the sacculus has to be elongated and two new cell poles have to be synthesized. The elongation is performed by a protein complex, the elongase that inserts disaccharidepentapeptide units at a limited number of discrete sites

  14. A cw 4-rod RFQ linac

    International Nuclear Information System (INIS)

    Fujisawa, Hiroshi

    1994-01-01

    A cw 4-rod RFQ linac system has been designed, constructed, and tested as an accelerator section of a MeV-class ion implanter system. The tank diameter is only 60 cm for 34 MHz operating frequency. An equally spaced arrangement of the RFQ electrode supporting plates is proved to be suitable for a low resonant frequency 4-rod RFQ structure. The RFQ electrode cross section is not circular but rectangular to make the handling and maintenance of the electrodes easier. The machining of the electrode is done three dimensionally. Second order corrections in the analyzing magnet of the LEBT (Low Energy Beam Transport) section assure a better transmission through and the matching to the RFQ. A new approach is introduced to measure the rf characteristics of the 4-rod RFQ. This method requires only a few capacitors and a network analyzer. Both the rf and thermal stability of the 4-rod RFQ are tested up to cw 50 kW. Beam experiments with several ions confirm the acceleration of beams to the goal energy of 83 keV/u. The ion beam intensities obtained at the RFQ output for He + , N 2+ , and C + are 32, 13, and 220 pμA, respectively. The measured beam transmissions of >80% agree with the PARMTEQ calculations. The ion implantation method also gives definitive information on the energies of an RFQ output beam. ((orig.))

  15. Dry Rod Consolidation Technology Project results

    International Nuclear Information System (INIS)

    Mullen, C.K.; Feldman, E.M; Vinjamuri, K.; Griebenow, B.L.; Lynch, R.J.; Arave, A.E.; Hill, R.C.

    1988-01-01

    The Dry Rod Consolidation Technology (DRCT) Project conducted at the Idaho National Engineering Laboratory (INEL), in 1987 demonstrated the technical feasibility of a dry horizontal fuel rod consolidation process. Fuel rods from Westinghouse 15 /times/ 15 pressurized water reactor (PWR) spent fuel assemblies were consolidated into canisters to achieve a 2:1 volume reduction ratio. The consolidation equipment was operated at an existing hot cell complex at the INEL. The equipment was specifically designed to interface with the existing facility fuel handling and operational capabilities and was instrumented to provide data collection for process technology research. During the operational phase, data were collected from observation of the consolidation process, fuel assembly handling, and fuel rod behavior and characteristics. Equipment performance was recorded and data measurements were compiled on crud and contamination generated and spread. Fuel assembly skeletons [non-fuel bearing components (NFBC)] were gamma scanned and analyzed for isotopic content and profile. The above data collection was enhanced by extensive photograph and video documentation. The loaded consolidation fuel canisters were utilized for a test of the Transnuclear, Inc. TN-24P dry storage cask with consolidated fuel. The NFBC material was stored for a future volume reduction demonstration project. 14 figs., 4 tabs

  16. Coupler for nuclear reactor absorber rods

    International Nuclear Information System (INIS)

    Kerz, K.

    1984-01-01

    A coupler is described for absorber rods being suspended during operation of nuclear reactors which includes plurality of actuating elements being movable for individually and jointly releasing the coupler, the movement of each of the actuating elements for releasing the coupler being independently controllable

  17. Lessons from feedback of safety operating experience for reactor physics

    International Nuclear Information System (INIS)

    Suchomel, J.; Rapavy, S.

    1999-01-01

    Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

  18. Analysis of buffering process of control rod hydraulic absorber

    International Nuclear Information System (INIS)

    Bao Jishi; Qin Benke; Bo Hanliang

    2011-01-01

    Control Rod Hydraulic Drive Mechanism(CRHDM) is a newly invented build-in control rod drive mechanism. Hydraulic absorber is the key part of this mechanism, and is used to cushion the control rod when the rod scrams. Thus, it prevents the control rod from being deformed and damaged. In this paper dynamics program ANSYS CFX is used to calculate all kinds of flow conditions in hydraulic absorber to obtain its hydraulic characteristics. Based on the flow resistance coefficients obtained from the simulation results, fluid mass and momentum equations were developed to get the trend of pressure change in the hydraulic cylinder and the displacement of the piston rod during the buffering process of the control rod. The results obtained in this paper indicate that the hydraulic absorber meets the design requirement. The work in this paper will be helpful for the design and optimization of the control rod hydraulic absorber. (author)

  19. ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

    2004-01-01

    An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

  20. Numerical analysis of the fusion of nuclear combustible rods under LOCA - type accidents

    International Nuclear Information System (INIS)

    Idelsohn, S.R.; Crivelli, L.A.

    1983-01-01

    The study of the melting of combustible rods is of great importance for the safety analysis of nuclear reactors. Due to the special characteristics of the problem, a sharp interface between the solid and liquid region does not exist, but appears a 'mushy' region in which the material is partially melted. The Finite Element Method is employed here, together with a regularized enthalpy formulation. Finally, the results obtained are presented and discussed. (Author) [pt

  1. Calculated GCFR fuel rod behavior for steady state and transient operation

    International Nuclear Information System (INIS)

    Resch, S.C.

    1981-01-01

    The Idaho National Engineering Laboratory (INEL) was contracted to review the Preliminary Safety Information Document (PSID) Amendment 10 for Gas-Cooled Fast Reactors (GCFR). As part of this effort the light water reactor codes, FRAPCON-1 and FRAP-T5 were converted to model GCFR fuel rod behavior. The conversion and application of these codes for GCFR analyses is the subject of this paper

  2. Experimental studies of the effect of rod spacing on burnout in a simulated rod bundle

    International Nuclear Information System (INIS)

    Lee, D.H.; Little, R.B.

    1962-08-01

    Tests on a dumb-bell shaped flow passage simulating the gap between rods in a fuel element indicated that burnout was not significantly affected by inter-rod gap in the range 0.032'' to 0.22''. Test conditions were: 960 p.s.i.a., 2 x 10 6 1b/ft 2 hr mass velocity, and 10% mean exit quality with vertical upflow of water. (author)

  3. Non-parametric order statistics method applied to uncertainty propagation in fuel rod calculations

    International Nuclear Information System (INIS)

    Arimescu, V.E.; Heins, L.

    2001-01-01

    Advances in modeling fuel rod behavior and accumulations of adequate experimental data have made possible the introduction of quantitative methods to estimate the uncertainty of predictions made with best-estimate fuel rod codes. The uncertainty range of the input variables is characterized by a truncated distribution which is typically a normal, lognormal, or uniform distribution. While the distribution for fabrication parameters is defined to cover the design or fabrication tolerances, the distribution of modeling parameters is inferred from the experimental database consisting of separate effects tests and global tests. The final step of the methodology uses a Monte Carlo type of random sampling of all relevant input variables and performs best-estimate code calculations to propagate these uncertainties in order to evaluate the uncertainty range of outputs of interest for design analysis, such as internal rod pressure and fuel centerline temperature. The statistical method underlying this Monte Carlo sampling is non-parametric order statistics, which is perfectly suited to evaluate quantiles of populations with unknown distribution. The application of this method is straightforward in the case of one single fuel rod, when a 95/95 statement is applicable: 'with a probability of 95% and confidence level of 95% the values of output of interest are below a certain value'. Therefore, the 0.95-quantile is estimated for the distribution of all possible values of one fuel rod with a statistical confidence of 95%. On the other hand, a more elaborate procedure is required if all the fuel rods in the core are being analyzed. In this case, the aim is to evaluate the following global statement: with 95% confidence level, the expected number of fuel rods which are not exceeding a certain value is all the fuel rods in the core except only a few fuel rods. In both cases, the thresholds determined by the analysis should be below the safety acceptable design limit. An indirect

  4. Investigation of control rod worth and nuclear end of life of BWR control rods

    International Nuclear Information System (INIS)

    Magnusson, Per

    2008-01-01

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of 10 B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% 10 B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in 10 B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming

  5. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  6. Low fluid level in pulse rod shock absorber

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1974-01-01

    On various occasions during pulse mode operation the shim and regulating control rods would drop when the pulse rod was withdrawn. Subsequent investigation traced the problem to the pulse rod shock absorber which was found to be low in hydraulic fluid. The results of the investigation, the corrective action taken, and a method for measuring the shock absorber fluid level are presented. (author)

  7. Summary of Skoda JS rod drop measurements analysis

    International Nuclear Information System (INIS)

    Svarny, J.; Krysl, V.

    1999-01-01

    A summary is presented of the Skoda JS rod drop reactivity measurements analysis provided during last two years based on control rod worth measurements by the outer ion chambers. Standard analysis based on comparisons of dynamics macrocode MOBY-DICK-SK and experimental data is extended to the 8-th group delayed neutron structure and new features of rod drop process are investigated. (author)

  8. Low fluid level in pulse rod shock absorber

    Energy Technology Data Exchange (ETDEWEB)

    Aderhold, H. C.

    1974-07-01

    On various occasions during pulse mode operation the shim and regulating control rods would drop when the pulse rod was withdrawn. Subsequent investigation traced the problem to the pulse rod shock absorber which was found to be low in hydraulic fluid. The results of the investigation, the corrective action taken, and a method for measuring the shock absorber fluid level are presented. (author)

  9. Preliminary nuclear design for test MOX Fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

    1997-10-01

    As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

  10. Single-phase convective heat transfer in rod bundles

    International Nuclear Information System (INIS)

    Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

    2008-01-01

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

  11. Single-phase convective heat transfer in rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

    2008-04-15

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

  12. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

  13. Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

    International Nuclear Information System (INIS)

    Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

    2012-01-01

    Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

  14. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  15. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  16. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    Pantoja C, R.

    2010-01-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  17. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    Jacquelin, Roland.

    1977-01-01

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container [fr

  18. Rebirth of a control rod at the Phenix power plant

    International Nuclear Information System (INIS)

    De Carvalho, Corinne; Vignau, Bernard; Masson, Marc

    2007-01-01

    This paper outlines the operations involved in cleaning the control rod for the complementary shutdown system in the Phenix Power Plant, the French sodium-cooled fast reactor. The Phenix reactor is controlled by six control rods and a complementary shutdown system. The latter comprises a control rod and a mechanism maintaining the rod in position by means of an electromagnet. The electromagnet is continuously supplied with power and holds the rod control assembly in position by magnetisation on a plane circular surface made from pure iron. The bearing capacity of the mechanism on the rod was initially 80 daN with a rod weight of 26.3 daN. This deteriorated progressively over time. The bearing surface of the rod and the electromagnet became contaminated with a deposit of sodium oxides and metallic particles, thus creating an air gap. This reached a figure of 36 daN in 2005 and was deemed not to be sufficient to prevent the rod from dropping at the wrong time during reactor operation. The Power Plant thus decided to replace the rod mechanism in the reactor in an initial phase, followed by the control rod itself. As the Phenix Power Plant had no spare control rods left, they initiated a 'salvage' plan, over two stages, for the rod removed from the reactor and placed in the fuel storage drum: - Inspection of the bearing surface of the rod by means of a borescope to check whether the rod could be salvaged, - A cleaning operation on the bearing face and checks on the bearing capacity of the rod. The operation is subject to very stringent requirements: the rod must not be taken out of the sodium to ensure that it can be reused in the reactor. The operation must thus take place in the fuel storage drum where there are no facilities for such an operation and where operating conditions are very hostile: high temperatures (the sodium in the fuel storage drum is at a temperature of 150 deg. C, high dose rate (3 mGy/h on the bearing surface) and the bearing surface is submerged

  19. Verification test of control rod system for HTR-10

    International Nuclear Information System (INIS)

    Zhou Huizhong; Diao Xingzhong; Huang Zhiyong; Cao Li; Yang Nianzu

    2002-01-01

    There are 10 sets of control rods and driving devices in 10 MW High Temperature Gas-cooled Test Reactor (HTR-10). The control rod system is the controlling and shutdown system of HTR-10, which is designed for reactor criticality, operation, and shutdown. In order to guarantee technical feasibility, a series of verification tests were performed, including room temperature test, thermal test, test after control rod system installed in HTR-10, and test of control rod system before HTR-10 first criticality. All the tests data showed that driving devices working well, control rods running smoothly up and down, random position settling well, and exactly position indicating

  20. Core design with respect to the safety concept

    International Nuclear Information System (INIS)

    Kollmar, W.

    1981-01-01

    In the present paper the following topics are dealt with: Principles of reactor core design and optimization, fuel management and safety concept for higher cycles and results of risk analyses (e.g. rod ejection, steam line break etc.) (RW)

  1. Photonic mesophases from cut rod rotators

    Energy Technology Data Exchange (ETDEWEB)

    Stelson, Angela C.; Liddell Watson, Chekesha M., E-mail: cml66@cornell.edu [Materials Science and Engineering, Cornell University, Ithaca, New York 14853 (United States); Avendano, Carlos [Chemical Engineering and Analytical Science, The University of Manchester, Manchester M13 9PL (United Kingdom)

    2016-01-14

    The photonic band properties of random rotator mesophases are calculated using supercell methods applied to cut rods on a hexagonal lattice. Inspired by the thermodynamic mesophase for anisotropic building blocks, we vary the shape factor of cut fraction for the randomly oriented basis. We find large, stable bandgaps with high gap isotropy in the inverted and direct structures as a function of cut fraction, dielectric contrast, and filling fraction. Bandgap sizes up to 34.5% are maximized at high dielectric contrast for rods separated in a matrix. The bandgaps open at dielectric contrasts as low as 2.0 for the transverse magnetic polarization and 2.25 for the transverse electric polarization. Additionally, the type of scattering that promotes the bandgap is correlated with the effect of disorder on bandgap size. Slow light properties are investigated in waveguide geometry and slowdown factors up to 5 × 10{sup 4} are found.

  2. Composites reinforcement by rods a SAS study

    CERN Document Server

    Urban, V; Pyckhout-Hintzen, W; Richter, D; Straube, E

    2002-01-01

    The mechanical properties of composites are governed by size, shape and dispersion degree of so-called reinforcing particles. Polymeric fillers based on thermodynamically driven microphase separation of block copolymers offer the opportunity to study a model system of controlled rod-like filler particles. We chose a triblock copolymer (PBPSPB) and carried out SAS measurements with both X-rays and neutrons, in order to characterize separately the hard phase and the cross-linked PB matrix. The properties of the material depend strongly on the way that stress is carried and transferred between the soft matrix and the hard fibers. The failure of the strain-amplification concept and the change of topological contributions to the free energy and scattering factor have to be addressed. In this respect the composite shows a similarity to a two-network system, i.e. interpenetrating rubber and rod-like filler networks. (orig.)

  3. [Rod of Asclepius. Symbol of medicine].

    Science.gov (United States)

    Young, Pablo; Finn, Bárbara C; Bruetman, Julio E; Cesaro Gelos, Jorge; Trimarchi, Hernán

    2013-09-01

    Symbolism is one of the most archaic forms of human thoughts. Symbol derives from the Latin word symbolum, and the latter from the Greek symbolon or symballo, which means "I coincide, I make matches". The Medicine symbol represents a whole series of historical and ethical values. Asclepius Rod with one serpent entwined, has traditionally been the symbol of scientific medicine. In a misconception that has lasted 500 years, the Caduceus of Hermes, entwined by two serpents and with two wings, has been considered the symbol of Medicine. However, the Caduceus is the current symbol of Commerce. Asclepius Rod and the Caduceus of Hermes represent two professions, Medicine and Commerce that, in ethical practice, should not be mixed. Physicians should be aware of their real emblem, its historical origin and meaning.

  4. Control rod drives for nuclear reactors

    International Nuclear Information System (INIS)

    Okada, Shige.

    1981-01-01

    Purpose: To enable the detection for the raptures of the bellows in control rod drives in LMFBR type reactor, by recycling seal gases inside the bellows and measuring the radioactivity in the recycling passage. Constitution: In the control drives, outer extension pipe is surrounded by the bellows, which is put between the cylindrical biological shieldings around the upper potion of an upper guide tube and the disk-like seal members provided at the lower flange of the outer extension pipe. Thus, the inside device of control rod is isolated from the coolants and the cover gases. The outer extension pipe is provided with a suction channel and a return channel. These channels are connected to a seal gas recycling pipeway, a pump and a radioactivity detector, where the seal gases in the bellows is recycling. If failures should occur in the bellows, cover gas leaks into the seal gas and recycles, whereby radioactivity is detected and alarmed. (J.P.N.)

  5. The lifetime of the control rod drives

    International Nuclear Information System (INIS)

    Avet, B.; Cauquelin, C.

    1989-01-01

    The lifetime of the control rod drives is studied. Their function is to take out or to pull in the control rods. The drive and the experiments carried out, are described. The analysis of the behaviour under operation, the drive inspections and surveyance, are also considered. The results are obtained from: the investigations performed on the fatigue strength of the 900 MW and 1300 MW drives, which allowed to deduce a low of wear and to identify the important aspects to be studied, the measurements of the dynamical stresses of mobile elements and a dynamical calculation model. The study leads to the conclusion that a probabilistic approach is needed for the fatigue damage analysis of some elements. Moreover, a systematic examination is also needed, to verify the agreement betwem the drives calculated aging values and the measured ones [fr

  6. Fuel rod failure detection method and system

    International Nuclear Information System (INIS)

    Assmann, H.; Janson, W.; Stehle, H.; Wahode, P.

    1975-01-01

    The inventor claims a method for the detection of a defective fuel rod cladding tube or of inleaked water in the cladding tube of a fuel rod in the fuel assembly of a pressurized-water reactor. The fuel assembly is not disassembled but examined as a whole. In the examination, the cladding tube is heated near one of its two end plugs, e.g. with an attached high-frequency inductor. The water contained in the cladding tube evaporates, and steam bubbles or a condensate are detected by the ultrasonic impulse-echo method. It is also possible to measure the delay of the temperature rise at the end plug or to determine the cooling energy required to keep the end plug temperature stable and thus to detect water ingression. (DG/AK) [de

  7. Passive cooling of control rod drive mechanisms

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Schwirian, R.E.

    1992-01-01

    A method and apparatus are provided for passively cooling the control rod drive mechanisms (CRDMs) in the reactor vessel of a nuclear power plant. Passive cooling is achieved by dispersing a plurality of chimneys within the CRDM array in positions where a control rod is not required. The chimneys induce convective air currents which cause ambient air from within the containment to flow over the CRDM coils. The air heated by the coils is guided into inlets in the chimneys by baffles. The chimney is insulated and extends through the seismic support platform and missile shield disposed above the closure head. A collar of adjustable height mates with plate elements formed at the distal end of the CRDM pressure housings by an interlocking arrangement so that the seismic support platform provides lateral restraint for the chimneys. (Author)

  8. Solitary waves on nonlinear elastic rods. I

    DEFF Research Database (Denmark)

    Sørensen, Mads Peter; Christiansen, Peter Leth; Lomdahl, P. S.

    1984-01-01

    Acoustic waves on elastic rods with circular cross section are governed by improved Boussinesq equations when transverse motion and nonlinearity in the elastic medium are taken into account. Solitary wave solutions to these equations have been found. The present paper treats the interaction betwe...... nonlinearity. The balance between dispersion and nonlinearity in the equation is investigated.......Acoustic waves on elastic rods with circular cross section are governed by improved Boussinesq equations when transverse motion and nonlinearity in the elastic medium are taken into account. Solitary wave solutions to these equations have been found. The present paper treats the interaction between...... the solitary waves numerically. It is demonstrated that the waves behave almost like solitons in agreement with the fact that the improved Boussinesq equations are nearly integrable. Thus three conservation theorems can be derived from the equations. A new subsonic quasibreather is found in the case of a cubic...

  9. SSYST. A code system to analyze LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1982-01-01

    SSYST (Safety SYSTem) is a modular system to analyze the behavior of light water reactor fuel rods and fuel rod simulators under accident conditions. It has been developed in close cooperation between Kernforschungszentrum Karlsruhe (KfK) and the Institut fuer Kerntechnik und Energiewandlung (IKE), University Stuttgart, under contract of Projekt Nukleare Sicherheit (PNS) at KfK. Although originally aimed at single rod analysis, features are available to calculate effects such as blockage ratios of bundles and wholes cores. A number of inpile and out-of-pile experiments were used to assess the system. Main differences versus codes like FRAP-T with similar applications are (1) an open-ended modular code organisation, (2) availability of modules of different sophistication levels for the same physical processes, and (3) a preference for simple models, wherever possible. The first feature makes SSYST a very flexible tool, easily adapted to changing requirements; the second enables the user to select computational models adequate to the significance of the physical process. This leads together with the third feature to short execution times. The analysis of transient rod behavior under LOCA boundary conditions e.g. takes 2 mins cpu-time (IBM-3033), so that extensive parametric studies become possible

  10. Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

    2010-01-01

    The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

  11. Evaluation of the rod ejection accident in Westinghouse Pressurized Water Reactors using spatial kinetics methods

    International Nuclear Information System (INIS)

    Risher, D.H. Jr.

    1975-01-01

    The consequences of a rod ejection accident are investigated in relation to the latest, high power density Westinghouse reactors. Limiting criteria are presented, based on experimental evidence, and if not exceeded these criteria will ensure that there will be no interference with core cooling capability, and radiation releases, if any, will be within the guidelines of 10CFR100. A basis is presented for the conservative selection of plant parameters to be used in the analysis, such that the analysis is applicable to a wide range of past, present, and future reactors. The calculational method employs a one-dimensional spatial kinetics computer code and a transient fuel heat transfer computer code to determine the hot spot fuel temperature versus time following a rod ejection. Using these computer codes, the most limiting hot channel factor (which does not cause the fuel damage limit criteria to be exceeded) has been determined as a function of the ejected rod worth. By this means, the limit criteria have been translated into ejected rod worths and hot channel factors which can be used effectively by the nuclear designer and safety analyst. The calculational method is shown to be conservative, compared to the results of a three-dimensional spatial kinetics analysis

  12. Application of pattern recognition techniques to the detection of the Phenix reactor control rods vibrations

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Deat, M.; Le Guillou, G.

    1979-01-01

    The incipient detection of control rods vibrations is very important for the safety of the operating plants. This detection can be achieved by an analysis of the peaks of the power spectrum density of the neutron noise. Pattern Recognition techniques were applied to detect the rod vibrations which occured at the fast breeder Phenix (250MWe). In the first part we give a description of the basic pattern which is used to characterize the behavior of the plant. The pattern is considered as column vector in n dimensional Euclidian space where the components are the samples of the power spectral density of the neutron noise. In the second part, a recursive learning procedure of the normal patterns which provides the mean and the variance of the estimates is described. In the third part the classification problem has been framed in terms of a partitioning procedure in n dimensional space which encloses regions corresponding to normal operations. This pattern recognition scheme was applied to the detection of rod vibrations with neutron data collected at the Phenix site before and after occurence of the vibrations. The analysis was carried out with a 42-dimensional measurement space. The learned pattern was estimated with 150 measurement vectors which correspond to the period without vibrations. The efficiency of the surveillance scheme is then demonstrated by processing separately 119 measurement vectors recorded during the rod vibration period

  13. Crossflow between subchannels in a 5 x 5 rod-bundle geometry

    Science.gov (United States)

    Lee, Jungjin; Park, Hyungmin

    2017-11-01

    In the present study, we experimentally investigate the single-phase (water as a working fluid) flow in a vertical 5 x 5 rod-bundle geometry using a particle image velociemtry, especially focusing on the crossflow phenomena between subchannels. This crossflow phenomena is very important in determining the performance and safety of nuclear power plant. To measure the flow behind the rod, it is made of FEP (Fluorinated Ethylene Propylene) to achieve the index matching. The ratio of pitch between rods and rod diameter is 1.4, and the considered Reynolds number based on a hydraulic diameter of a channel and an axial bulk velocity is 10000. Also, the typical grid spacer is installed periodically along the streamwise direction. Depending on the location of subchannel (e.g., distance to the side wall or grid spacer), the flow (turbulence) statistics show large variations that will be discussed in detail. Furthermore, we will suggest a modified crossflow model that can explain the varying crossflow phenomena more clearly. Supported by NRF Grant (NRF-2016M2B2A9A02945068) of the Korean government.

  14. Grid for nuclear fuel rod assembly

    International Nuclear Information System (INIS)

    Brayman, K.W.; George, D.K.; Rawlings, J.C.; Dix, G.E.

    1976-01-01

    A grid is described for placing a least four corner fuel rods in a tubular flow channel of a nuclear reactor. It includes a bearer component composed of four side strips joined by four corner strips so as to form a rigid unit structure, each side strip having an L-shaped piece adjacent at each of its ends to a lug of each L-shaped piece extending to the adjacent end of its associated side strip [fr

  15. Hydrodynamic behavior of a bare rod bundle

    International Nuclear Information System (INIS)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers

  16. Experience with a fuel rod enrichment scanner

    International Nuclear Information System (INIS)

    Kubik, R.N.; Pettus, W.G.

    1975-01-01

    This enrichment scanner views all fuel rods produced at B and W's Commercial Nuclear Fuel Plant. The scanner design is derived from the PAPAS System reported by R. A. Forster, H. D. Menlove, and their associates at Los Alamos. The spatial resolution of the system and smoothing of the data are discussed in detail. The cost-effectiveness of multi-detector versus single detector scanners of this general design is also discussed

  17. A light ion four rod RFQ injector

    International Nuclear Information System (INIS)

    Schempp, A.; Ferch, M.; Klein, H.

    1987-01-01

    The four-rod RFQ has been developed in Frankfurt as an alternative solution for ion injectors. A 202 MHz resonator has been built with design parameters taken from the HERA injector (18keV-750keV, 20mA H - ). Properties of this structure are described and applications as light ion accelerator for particles from an EBIS ion source are discussed

  18. Towards Viscoplastic Constitutive Models for Cosserat Rods

    OpenAIRE

    Dörlich Vanessa; Linn Joachim; Scheffer Tobias; Diebels Stefan

    2016-01-01

    Flexible, slender structures like cables, hoses or wires can be described by the geometrically exact Cosserat rod theory. Due to their complex multilayer structure, consisting of various materials, viscoplastic behavior has to be expected for cables under load. Classical experiments like uniaxial tension, torsion or three-point bending already show that the behavior of e.g. electric cables is viscoplastic. A suitable constitutive law for the observed load case is crucial for a realistic simul...

  19. On the perfect hexagonal packing of rods

    International Nuclear Information System (INIS)

    Starostin, E L

    2006-01-01

    In most cases the hexagonal packing of fibrous structures or rods extremizes the energy of interaction between strands. If the strands are not straight, then it is still possible to form a perfect hexatic bundle. Conditions under which the perfect hexagonal packing of curved tubular structures may exist are formulated. Particular attention is given to closed or cycled arrangements of the rods like in the DNA toroids and spools. The closure or return constraints of the bundle result in an allowable group of automorphisms of the cross-sectional hexagonal lattice. The structure of this group is explored. Examples of open helical-like and closed toroidal-like bundles are presented. An expression for the elastic energy of a perfectly packed bundle of thin elastic rods is derived. The energy accounts for both the bending and torsional stiffnesses of the rods. It is shown that equilibria of the bundle correspond to solutions of a variational problem formulated for the curve representing the axis of the bundle. The functional involves a function of the squared curvature under the constraints on the total torsion and the length. The Euler-Lagrange equations are obtained in terms of curvature and torsion and due to the existence of the first integrals the problem is reduced to the quadrature. The three-dimensional shape of the bundle may be readily reconstructed by integration of the Ilyukhin-type equations in special cylindrical coordinates. The results are of universal nature and are applicable to various fibrous structures, in particular, to intramolecular liquid crystals formed by DNA condensed in toroids or packed inside the viral capsids

  20. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  1. Low reactivity penalty burnable poison rods

    International Nuclear Information System (INIS)

    1978-01-01

    A nuclear reactor burnable poison rod is described which consists of an elongated tubular sheath enclosing a neutron absorbing material which, at least during reactor operation, also encloses a neutron moderating material. The excess reactivity existing at the beginning of core life is compensated for by the depletion of the burnable poison throughout the life of the core, so that the life of the core is extended. (UK)

  2. Cone dystrophy with "supernormal" rod ERG: psychophysical testing shows comparable rod and cone temporal sensitivity losses with no gain in rod function.

    Science.gov (United States)

    Stockman, Andrew; Henning, G Bruce; Michaelides, Michel; Moore, Anthony T; Webster, Andrew R; Cammack, Jocelyn; Ripamonti, Caterina

    2014-02-10

    We report a psychophysical investigation of 5 observers with the retinal disorder "cone dystrophy with supernormal rod ERG," caused by mutations in the gene KCNV2 that encodes a voltage-gated potassium channel found in rod and cone photoreceptors. We compared losses for rod- and for cone-mediated vision to further investigate the disorder and to assess whether the supernormal ERG is associated with any visual benefit. L-cone, S-cone, and rod temporal acuity (critical flicker fusion frequency) were measured as a function of target irradiance; L-cone temporal contrast sensitivity was measured as a function of temporal frequency. Temporal acuity measures revealed that losses for vision mediated by rods, S-cones, and L-cones are roughly equivalent. Further, the gain in rod function implied by the supernormal ERG provides no apparent benefit to near-threshold rod-mediated visual performance. The L-cone temporal contrast sensitivity function in affected observers was similar in shape to the mean normal function but only after the mean function was compressed by halving the logarithmic sensitivities. The name of this disorder is potentially misleading because the comparable losses found across rod and cone vision suggest that the disorder is a generalized cone-rod dystrophy. Temporal acuity and temporal contrast sensitivity measures are broadly consistent with the defect in the voltage-gated potassium channel producing a nonlinear distortion of the photoreceptor response but after otherwise normal transduction processes.

  3. Installing and detaching apparatus for a control rod drive mechanism

    International Nuclear Information System (INIS)

    Akimoto, Seiichi; Watanabe, Mitsuhiro; Yoshida, Tomiharu; Sugaya, Jun-ichi; Saito, Takashi.

    1976-01-01

    Object: To facilitate maintenance and repair of a control rod drive mechanism. Structure: The apparatus comprises a means moving in a moving direction of a control rod within a reactor vessel, said moving means having a housing mounted thereon, a means mounted on the reactor vessel to release a connection between a control rod drive mechanism connected to the control rod and the control rod, and a means for mounting and removing a fixing means which connects the reactor vessel to the control rod drive means. With this arrangement, cooling water of high radioactivity level may not be leaked outside to thereby notably reduce dangerousness of exposure and materially cut time required for mounting and removing the control rod drive mechanism. (Ohara, T.)

  4. Analytical prediction of turbulent friction factor for a rod bundle

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Park, Joo Hwan

    2011-01-01

    An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels.

  5. Noise problems in PFR safety system power supplies

    Energy Technology Data Exchange (ETDEWEB)

    Hodgson, D

    1978-08-15

    Reactor power on PFR is controlled by eleven neutron absorbing rods. They are held vertically above the core by electromagnets which, in turn, are mounted on motorised lead screws. Five of the rods are used as power control elements; five are used as shut off rods and the eleventh is a safety rod, positioned vertically above the centre of the reactor core. All rods fall under gravity into the reactor when the electro magnets are de-energised. In addition the safety rod can be driven into the reactor by a pneumatic piston and cylinder mechanism. This provides some degree of diversity from the common design of mechanism used by the control and shut off rods. Also being in the center it can be armed and ready for use when the above core shield is rotated. The automatic protective system for the plant provides the electro magnet currents and hence the reactor trip signals via two separate and diverse safety systems, namely a ''Relay System'' and a ''Solid State System.'' The two systems are completely separate and independent. The overall probability of failure of the complete system to release at least 3 rods when called upon to do so is less than 10/sup -6/.

  6. Magnetic switch for reactor control rod

    International Nuclear Information System (INIS)

    Germer, J.H.

    1986-01-01

    This patent describes a control rod system for a nuclear reactor utilizing an electromagnetic grapple mechanism for holding and releasing a control rod, the improvement comprising a magnetic reed switch assembly having a Curie-point magnetic shunt and responsive to reactor coolant temperature for short circuiting the electromagnetic grapple mechanism causing release of the control rod when the coolant temperature reaches the Curie-point of the magnetic shunt. The magnetic reed switch assembly includes a: a permanent magnet, a pair of magnetic pole pieces located at and in contact with opposite ends of the permanent magnet, the Curie-point magnetic shunt being positioned adjacent the permanent magnet and in contact with the pair of magnetic pole pieces, and a reed switch positioned intermediate the pole pieces and provided with a pair of ferromagnetic reeds, a nonmagnetic enclosure around the reeds, a first of the reeds being secured at one end to a first of the pair of pole pieces, a second of the reeds having one end extending into and secured to a hollow member positioned in and extending through a second of the pair of pole pieces, the one end of the second of the reeds secured to a condector adapted to be connected to the electromagnetic grapple mechanism

  7. LOFT fuel rod surface temperature measurement testing

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.; Solbrig, C.W.

    1978-01-01

    Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has been done to characterize the thermocouple design. Thermal cycling and corrosion testing of the thermocouple weld design have provided an expected lifetime of 6000 hours when exposed to reactor coolant conditions of 620 K and 15.9 MPa and to sixteen thermal cycles with an initial temperature of 480 K and peak temperatures ranging from 870 to 1200K. Departure from nucleate boiling (DNB) tests have indicated a DNB penalty (5 to 28% lower) during steady state operation and negligible effects during LOCA blowdown caused by the LOFT fuel rod surface thermocouple arrangement. Experience with the thermocouple design in Power Burst Facility (PBF) and LOFT nonnuclear blowdown testing has been quite satisfactory. Tests discussed here were conducted using both stainless steel and zircaloy-clad electrically heated rod in the LOFT Test Support Facility (LTSF) blowdown simulation loop

  8. Removable control rod drive shaft guide

    International Nuclear Information System (INIS)

    Ales, M.W.; Brown, S.K.; Dixon, L.D.

    1988-01-01

    A removable control rod drive shaft guide is described for a control rod ''guide'' structure card, comprising: a. a substantially annular shaped main body portion having a central axial bore for receiving a control rod drive shaft and an upper exterior groove for receiving removal tooling; b. the main body portion having a reduced outer diameter at its lower section; c. a shoulder portion integral with the main body portion for supporting the main body portion on the guide structure card; d. the shoulder portion having a substantially radial bore and the reduced outer diameter lower section having a slot in alignment with the radial bore; e. a locking arm ''pivotaly'' mounted in the radial bore which protrudes into the slot and is movable between a first normal locking position for engaging the guide structure card and a second release position; f. a spring received within a second axial bore in the main body portion and biased against the locking arm for urging and locking arm into the first normal locking position; and g. a release tab at one end of the locking arm for moving the locking arm into the second release position

  9. Modeling the Flexural Rigidity of Rod Photoreceptors

    Science.gov (United States)

    Haeri, Mohammad; Knox, Barry E.; Ahmadi, Aphrodite

    2013-01-01

    In vertebrate eyes, the rod photoreceptor has a modified cilium with an extended cylindrical structure specialized for phototransduction called the outer segment (OS). The OS has numerous stacked membrane disks and can bend or break when subjected to mechanical forces. The OS exhibits axial structural variation, with extended bands composed of a few hundred membrane disks whose thickness is diurnally modulated. Using high-resolution confocal microscopy, we have observed OS flexing and disruption in live transgenic Xenopus rods. Based on the experimental observations, we introduce a coarse-grained model of OS mechanical rigidity using elasticity theory, representing the axial OS banding explicitly via a spring-bead model. We calculate a bending stiffness of ∼105 nN⋅μm2, which is seven orders-of-magnitude larger than that of typical cilia and flagella. This bending stiffness has a quadratic relation to OS radius, so that thinner OS have lower fragility. Furthermore, we find that increasing the spatial frequency of axial OS banding decreases OS rigidity, reducing its fragility. Moreover, the model predicts a tendency for OS to break in bands with higher spring number density, analogous to the experimental observation that transgenic rods tended to break preferentially in bands of high fluorescence. We discuss how pathological alterations of disk membrane properties by mutant proteins may lead to increased OS rigidity and thus increased breakage, ultimately contributing to retinal degeneration. PMID:23442852

  10. Incorporation of squalene into rod outer segments

    International Nuclear Information System (INIS)

    Keller, R.K.; Fliesler, S.J.

    1990-01-01

    We have reported previously that squalene is the major radiolabeled nonsaponifiable lipid product derived from [ 3 H]acetate in short term incubations of frog retinas. In the present study, we demonstrate that newly synthesized squalene is incorporated into rod outer segments under similar in vitro conditions. We show further that squalene is an endogenous constituent of frog rod outer segment membranes; its concentration is approximately 9.5 nmol/mumol of phospholipid or about 9% of the level of cholesterol. Pulse-chase experiments with radiolabeled precursors revealed no metabolism of outer segment squalene to sterols in up to 20 h of chase. Taken together with our previous absolute rate studies, these results suggest that most, if not all, of the squalene synthesized by the frog retina is transported to rod outer segments. Synthesis of protein is not required for squalene transport since puromycin had no effect on squalene incorporation into outer segments. Conversely, inhibition of isoprenoid synthesis with mevinolin had no effect on the incorporation of opsin into the outer segment. These latter results support the conclusion that the de novo synthesis and subsequent intracellular trafficking of opsin and isoprenoid lipids destined for the outer segment occur via independent mechanisms

  11. Position control device for a control rod

    International Nuclear Information System (INIS)

    Ono, Takehiko; Kusaka, Shuji.

    1976-01-01

    Purpose: To reliably prevent dangerous operation in the control of the position of the control rod by checking for abnormal pulse motor coil excitation voltage and, at the time of occurrence of abnormality, immediately holding the control rod stationary lest it should be moved to an unsafe position, this being accomplished excitation from a compensating excitation system. Constitution: In an FBR reactor, a circuit for memorizing the correct output states of individual drive signals at arbitrary instants and consequtively producing the memorized results is provided, and the output of the circuit and the actual drive signal are compared at all times to discriminate whether the drive signal being compared is normal or not. When the actual drive signal is abnormal, a series signal varying after a predetermined pattern is shifted to enable replacement of the actual drive signal, so that irrespective of whether the problem drive signal is ''on'' or ''off'', a drive signal of the correct pattern may be supplied to the pulse motor to hold the control rod and prevent it from being moved toward the dangerous side due to its own weight or other causes. (Horiuchi, T.)

  12. Control rod guide tube cleaning device

    International Nuclear Information System (INIS)

    Tsuji, Tadashi; Shiota, Yoshiaki.

    1990-01-01

    Since there was no exclusive device for cleaning control rods, no effective cleaning could not be conducted and there was a possibility that obstacles may not be recovered. Then, there are disposed a first pump for supplying pressurized water, a spray nozzle for forming a swirling flow in a control rod guide tube, a second pump for pressurizing water introduced by a sucking pipeline and a collecting device for recovering obstacles intruding to water from the second pump. The pressurized water supplied from the first pump is introduced to a head passing through a blowing pipe and jetted from the spray nozzle to the control rod guide tube. In this case, a swirling stream occurs and obstacles in the control guide tube are mixed into water. The water containing the obstacles passes from the sucking port through a pipeline, introduced to the second pump and recovered to the collecting device. Since there is no water staying portion upon cleaning operation, the obstacles accumulating over the entire region of the bottom of the guide tube can be recovered reliably and efficiently. (N.H.)

  13. Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

    International Nuclear Information System (INIS)

    Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

    1982-01-01

    The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

  14. Underwater characterization of control rods for waste disposal using SMOPY

    Energy Technology Data Exchange (ETDEWEB)

    Gallozzi-Ulmann, A.; Couturier, P.; Amgarou, K.; Rothan, D.; Menaa, N. [CANBERRA France,1 rue des Herons, 78182 ST Quentin Yvelines Cedex (France); Chard, P. [CANBERRA UK, Lower Dunbeath House, Forss Business Park, Thurso, Caithness KW14 7UZ (United Kingdom)

    2015-07-01

    Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performing radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were

  15. Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

    International Nuclear Information System (INIS)

    Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

    1999-10-01

    The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

  16. Relation of fuel rod service parameters and design requirements to produced fuel rod and their components

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.

    1999-01-01

    Based on the presented material it is possible to state that there is a very close link between the fuel operational parameters and the requirements for its design and production process. The required performance and life-time of a fuel rod can be only assured by the correctly selected design and process solutions. The economical aspect of this problem is significantly depend on the commercial feasibility of a particular selected solution with the provision of an automated and comparative by inexpensive production of a fuel rod and its components. The operational conditions are also important for the life time of the fuel rods. If there are no special measures for the mitigation of the certain operation conditions the leakage of fuel elements can take place before the planned time. (authors)

  17. Influence of initial conditions on rod behaviour during boiling crisis phase following a reactivity initiated accident

    International Nuclear Information System (INIS)

    Georgenthum, V.; Sugiyama, T.

    2010-01-01

    In the frame of their research programs on high burn-up fuel safety, the French Institute for Radioprotection and Nuclear Safety (IRSN) and the Japan Atomic Energy Agency (JAEA) performed a large set of tests devoted to the study of PWR fuel rod behavior during Reactivity Initiated Accident (RIA) respectively in the CABRI reactor and in the NSRR reactor. The reactor test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed until now with sodium coolant at 280 Celsius degrees and 3 bar. In the NSRR reactor most of the tests were performed with stagnant water at 20 C. degrees and atmospheric pressure but recently a new high temperature high pressure capsule has been developed which allows to performed tests at up to 280 Celsius degrees and 70 bar. The paper discusses the influence of test conditions on rod behaviour during boiling phase, based on tests results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad straining and possible ballooning. The analysis of the different test conditions makes it possible to discriminate the influence of initial conditions on the different phases of the transient and is useful for modelling and code development. (authors)

  18. Shielding device for control rod in nuclear reactor

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo; Tomatsu, Tsutomu.

    1995-01-01

    The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

  19. Temperature measurement in cans of fuel rods and fuel rod simulators

    International Nuclear Information System (INIS)

    Tschoeke, H.; Moeller, R.

    1977-01-01

    On the sodium-cooled 19-rod cluster model for the SNR 300 the can wall temperature distributions of the non-uniformly cooled rods were measured with thermocouples mounted in outer grooves in the peripheral zone, permitting, in connection with Ni solder, a practically undisturbed measurement. For a more exact determination of the local surface temperature a calibration method, the so-called double-wall method, was developed and applied. The description of this calibration method and the experimental results achieved until now are presented. (orig./RW) [de

  20. Study and modeling of fluctuating fluid forces exerted on fuel rods in pressurized water reactors

    International Nuclear Information System (INIS)

    Bhattacharjee, Saptarshi

    2016-01-01

    Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in the fuel rods. Due to friction, wear occurs at the contact locations between the spacer grid and the fuel rod. This could compromise the first safety barrier of the nuclear reactor by damaging the fuel rod cladding. In order to ensure the integrity of the cladding, it is necessary to know the random fluctuating forces acting on the rods. However, the spectra for these fluid forces are not well known. The goal of this PhD thesis was to use simple geometrical elements to check the reproducibility of realistic pressurized water reactor spacer grids. As a first step, large eddy simulations were performed on a concentric annular pipe for different mesh refinements using the CFD code Trio CFD (previously Trio U) developed by CEA. A mesh sensitivity study was performed to obtain an acceptable mesh for reproducing standard literature results. This information on mesh resolution was used when carrying out simulations using various geometric obstacles inside the pipe, namely, mixing vanes, circular spacer grid and a combination of square spacer grid with mixing vanes. The last of the three configurations is the closest to a realistic PWR fuel assembly. Structured mesh was generated for the annular pipe case and circular grid case. An innovative hybrid mesh was used for the two remaining cases of the mixing vanes and the square grid: keeping unstructured mesh around the obstacles and structured mesh in the rest of the domain. The inner wall of the domain was representative of the fuel rod cladding. Both hydraulic and wall pressure characteristics were analyzed for each case. The results for the square grid case were found to be an approximate combination of the mixing vane case and circular grid case. Simulation results were compared with experiments performed at CEA Cadarache. Some preliminary comparisons were also made with classical semi-empirical models. (author) [fr