Results 1 - 10 of 18531
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[en] The purpose of this work is to establish the proficiency testing schemes of a fuel assembly for nonstandard test method case. As the nuclear regulatory guide, 'the testing and inspections to be performed to verify the design characteristics of the fuel system components, including clad integrity, dimensions, fuel enrichment, burnable poison concentration, absorber composition, and characteristics of the fuel, absorber, and poison pellets, should be described. ∼'. In this guide, the fuel assembly test method is as that, the lateral and axial stiffness, lateral vibration, lateral and axial impact and the rotational stiffness test. These method cases are very important for the license service and providing some input data for the accident analysis model of FA. Therefore, all of these tests have to be executed as the authorized standard, for example, Korea Laboratory Accreditation Scheme (KOLAS). Unfortunately, the performance tests of a FA did not certified by the KOLAS. In order to receive the authorized test scheme, the proficiency testing schemes is most important item. For non-standard test case, the most of these tests be normally executed through the inter-laboratory comparisons. However, there is no standard, no certified reference material (CRM) for pressurized water reactor (PWR) fuel assembly. In this case, the most important point is that how to verify the validity of the performance test method of a fuel. Therefore, the inter-personnel testing scheme is proposed for this. For the proficiency testing of a fuel assembly performance test, the lateral bending test of a fuel assembly (FA) is executed using FAMeCT. The FAMeCT is a tester of a versatile function for a mechanical characterization of an actual size FA. Because of the absence of the CRM, the t-test method was selected. Null and alternative hypotheses were assumed and then t-value was evaluated as these hypotheses
[en] A performance analysis of the loss-of-fluid test (LOFT) external accelerometer measurements is presented. Along with complete descriptions of test programs that have been conducted, specific sources of measurement uncertainty are identified, quantified, and combined to provide an assessment of the ability of this measurement to satisfy the requirement for measurement of structural acceleration
[en] Light water reactors operators have experienced a number of occurrences of improper performance by safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) recommended that programs be developed and completed which would reevaluate the performance capabilities of BWR safety and relief valves. This report provides the results of the review of these programs and their results by the NRC and their consultant, EG and G Idaho, Inc. Specifically, this report has examined the response of the Licensee (Washington Public Power Supply System) for the Nuclear Plant No. 2 to the requirements of NUREG-0578 and subsequently NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CFR-50 have been met
[en] DVI and FD are our own design features adopted into APR1400. This study focuses on improving their performance for the APR+. From a preliminary study, conceptual design for DVI+ has been developed. Various separate effect tests were performed to evaluate the performance of the design concept and to develop a basic design. For DVI+ design, tests for DVI duct performance and vibration characteristics were performed. For FD+ design, sensitivity tests for major design parameters were performed. This study also includes a development of design for the reactor flow distribution test as well as FD+ overall performance test
[en] Descriptions are given of the Fort St. Vrain Dew Point Moisture Monitor (DPMM) System; the bases for the DPMM system response time requirements for safety related functions at the required reactor operating conditions; the results and evaluation of recent testing which measured the performance of the current system at simulated operating conditions; predicted response times for reactor power operation from 0 to 100 percent and a modification to provide improved response times for low-load and plant start-up conditions
[en] Test results and evaluations are presented which were obtained subsequent to the installation of dew point moisture monitor (DPMM) bypass valve actuators and sample flow controllers in the Fort St. Vrain HTGR. New sample loop filters were installed to obtain all test data for clean filter operation. Testing was accomplished by supplying primary coolant gas from the PCRV through the DPMM system to the reactor building ventilation system. This allows measurements to be taken independent of circulator operation. Evaluation of the results of the current tests indicates that the modified DPMM system performs satisfactorily and meets all response time criteria. Based on the data collected, an evaluation was made of the performance of the system with higher than usual sample loop resistance. Minimum sample flow rates and bypass valve positions that ensure acceptable response times are met were determined.
[en] The performance of the water leak detection system on PFR has been studied by injecting known quantities of hydrogen and water into the steam generator units. The results confirmed the performance predicted from laboratory and rig development work
[en] The purpose of the tests described is to show that the dynamic performance of the Fort St. Vrain helium circulator auxiliary systems satisfies all the guidelines and criteria established and agreed to by Public Service Company of Colorado (PSC), Proto-Power, and General Atomic Company (GA). Specifically, it is shown that transfers to and from backup bearing water and helium purification system transients do not cause any circulator trips. Furthermore, at PSC's request, in an effort to resolve any NFSC questions concerning these systems, the satisfactory repeatability of their dynamic performance is shown beyond any doubt.
[en] In a Sodium-cooled Fast Reactor (SFR), liquid sodium is subject to the formation of impurities by its high chemical reactivity with so many elements and common compounds used in nuclear reactor construction materials. The impurities are mainly in the form of hydrides, oxides, metallic compounds, metallic and carbon particles, which originate primarily from steam generator corrosion, moisture from system component surface, and leakage of air into the system. These are finally deposited in the form of the crystallization of sodium hydride (NaH) or sodium oxide (Na2O) at the cold points of the circuit, which may lead to the clogging of the narrowed sections or may damage the pump. Therefore, it is important to research purifying performance of a cold trap. Up to now, many studies for cold traps have been accomplished but the studies for their performance are still under execution. KAERI secured a design technology for a new high-capacity cold trap through a technical cooperation with Kawasaki in Japan. It will be used in Sodium integral effect Test Loop for safety simulation and Assessment (STELLA-1) to purify the sodium after performance test in a instruments performance test loop