Download or read book Analysis of a Small break Loss of coolant Accident in a Pressurised Water Reactor written by Nikos C. Markatos and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Analysis of a Small break Loss of coolant Accident in a Pressurised Water Reactor by the Laura Version of Phoenics written by Stephen M. Rawnsley and published by . This book was released on 1982 with total page 78 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Analysis of a Four Inch Small break Loss of coolant Accident in a Westinghouse Pressurized Water Reactor Using TRAC PF1 MOD1 written by Kimberley I. R. Knippel and published by . This book was released on 1988 with total page 324 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Analysis of a Small Break Loss of coolant Accident of Pressurized Water Reactor by APROS written by and published by . This book was released on 1995 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

Download or read book Relap5 analysis of a small break loss of coolant accident without emergency core cooling in a pressurized water reactor written by P. Kuan and published by . This book was released on 1984 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: A relap5 thermal-hydraulics model based on the design of the three mile island unit-2 nuclear reactor was used to calculate the pressurized water reactor behavior during an hypothesized small-break loss-of-coolant accident without emergency core cooling. the calculation was performed to help plan the united states nuclear regulatory commission's internationally sponsored severe fuel damage test program in the power burst facility. both the primary system and the secondary sides of the steam generators were modeled. the calculation gave fluid conditions and structural temperatures throughout the system. the results indicated that 2 h after the pipe break: the upper half of the fuel bundles would be partially liquefied by molten zircaloy; the thin structures (c-tubes and split tubes) inside the guide tubes and the core barrel would be partially melted, while other structures inside the reactor vessel would likely remain intact; the hydrogen generated from the oxidation of zircaloy by steam would fill the upper plenum; and the temperature of the primary system pressure boundary would remain low enough that no additional pipe failure would be expected to occur. the calculation also showed that the oxidation rate was limited by the supply of steam during the rapid oxidation of the zircaloy cladding.

Download or read book Guidebook to Light Water Reactor Safety Analysis written by P. B. Abramson and published by CRC Press. This book was released on 1985-01-01 with total page 418 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Guidebook to Light Water Reactor Safety Analysis brings together government and expert researchers entrusted with maintaining the safety of reactors, preventing incidents, and for creating the guidelines for responding appropriately to emergency situations. It includes an overview presented by the U.S. Nuclear Regulatory Commission. One of the most relevant compendiums of its time, it's a volume of both historical and scientific significance and well worth the consideration of those currently involved with maintaining reactor safety..

Download or read book Heat Transfer During a Small break Loss of coolant Accident in a Pressurised Water Reactor written by Nikos C. Markatos and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Break Spectrum Analyses for Small Break Loss of Coolant Accidents in a RESAR 3S Plant written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs.

Download or read book Effect of Reactor Coolant Pumps Following a Small Break in a Pressurized Water Reactor written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the optimum mode of pump operation is design dependent. In terms of primary system mass depletion, the case with no pump trip was preferable for the W plant, whereas an early pump trip was preferable for the B and W plant. When the pumps were not operating in the W plant, the loop seals plugged with liquid, leading to a pressure buildup in the upper plenum and, consequently, a high liquid flow through the break. The vent valves in the B and W plant mitigated the consequences of the loop seals plugging; the effect was enough to favor an early pump trip.

Download or read book Investigation of Small Break Loss of coolant Phenomena in a Small Scale Nonnuclear Test Facility PWR written by and published by . This book was released on 1980 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR.

Download or read book Best estimate Analysis of a Loss of coolant Accident in a Four loop US PWR Using TRAC PD2 written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950°K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated.

Download or read book Generic Assessment of Delayed Reactor Coolant Pump Trip During Small Break Loss of coolant Accidents in Pressurized Water Reactors written by Brian Sheron and published by . This book was released on 1979 with total page 80 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book An Analysis of the Loss of coolant Accident in a Pressurized water Nuclear Reactor written by Alan Edward Ladieu and published by . This book was released on 1968 with total page 252 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Small Break Loss of Coolant Accident Analyses in Light Water Reactors written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1981 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Proceedings of the U S Nuclear Regulatory Commission Water Reactor Safety Research Information Meeting written by and published by . This book was released on 1982 with total page 770 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Estimating Loss of Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.

Download or read book Anticipated and Abnormal Plant Transients in Light Water Reactors written by Pamela Lassahn and published by Springer Science & Business Media. This book was released on 2013-11-11 with total page 725 pages. Available in PDF, EPUB and Kindle. Book excerpt: Over the last 30 years, reactor safety technology has evolved not so much from a need to recover from accidents or incidents, but primarily from many groups in the nuclear community asking hypo thetical, searching (what if) ~uestions. This ~uestioning has indeed paid off in establishing preventive measures for many types of events and potential accidents. Conditions, such as reactivity excursions, large break, loss of coolant, core melt, and contain ment integrity loss, to name a few, were all at one time topics of protracted discussions on hypothesized events. Historically, many of these have become multiyear, large-scale research programs aimed at resolving the "what ifs. " For the topic of anticipated and abnormal plant transients, how ever, the searching ~uestions and the research were not so prolific until the mid-1970s. At that time, probabilistic risk methodolo gies began to tell us we should change our emphasis in reactor safety research and development and focus more on small pipe breaks and plant transients. Three Mile Island punctuated that message in 1979. The plant transient topic area is a multidisciplinary subject involving not only the nuclear, fluid flow, and heat transfer technologies, but also the synergistics of these with the reactor control systems, the safety s;,"stems, operator actions, maintenance and even management and the economic considerations of a given plant.