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LOFT Monthly Progress Report for February 1980

Description: The significant event of February was the on-schedule conduct of Test L3-2 on February 7. This was the second LOFT small break test with nuclear heat. It simulated a break of a one-inch pipe in a large commercial plant, whereas Test L3-1 had simulated a break of a four-inch pipe. For Test L3-2, the reactor plant and emergency core cooling system appeared to function as expected, although preliminary data evaluation indicates a higher break flow than expected, with a correspondingly greater depressurization. As the month ended, data evaluation was continuing. During February, Nuclear Regulatory Commission guidance was received that would require Tests L3-5 and L3-6 to use nuclear heat. Previously these tests, the next planned tests, had been designed as nonnuclear tests with and without operating coolant pumps. This revised guidance will require a replanning of the entire program schedule for better facility use. At the end of the month, replanning was underway. Costs for February are right on budget, although manpower levels are somewhat greater than budget. This latter variance results from an intentional manpower-material interchange.
Date: March 1, 1980
Creator: Kaufman, N. C.
Partner: UNT Libraries Government Documents Department

Experiment data report for 1-1/2-loop semiscale system isothermal tests 1008 and 1010

Description: Recorded test data are presented for Tests 1008 and 1010 of the isothermal portion of the Semiscale Blowdown and Emergency Core Cooling (ECC) Project. These data represent the results of the first isothermal tests utilizing an enlarged downcomer gap (1.0 inch) in the pressure vessel to investigate general system hydraulic response both with and without simulated ECC injection. The data, presented in the form of composite graphs in engineering units, have been analyzed only to the extent necessary to assume that they are reasonable and correct. (auth)
Date: March 1, 1974
Creator: Alder, R.S.; Feldman, E.M. & Pinson, P.A.
Partner: UNT Libraries Government Documents Department

Phenomenological and mathematical modeling of a high pressure steam driven jet injector. Part 2

Description: An injector is a particular type of jet pump which uses condensable vapor to entrain a liquid and discharge against a pressure higher than either motive or suction pressures. The injector has no moving parts and requires no external power supply nor any complex control system. Thus, the injector is particularly suited for emergency core cooling operations. A detailed survey has indicated that various injector designs are available for operating pressures below 250 psig. However, the design of these injectors from the viewpoint of a basic understanding of heat and mass transfer processes has not been well developed. A critical review of the models showed serious discrepancies between the analytical models and the experimental observations. The discrepancies evolved from the neglect of non-equilibrium aspects of the flow. The origin of the non-equilibrium aspects can be traced to the extremely small time scales governing the flow in the injector. Thus, time scales of the order of 10{sup {minus}2} seconds are involved in the injector, accompanied by mass, momentum, and heat transfer rates of orders of magnitude higher than that observed in conventional two-phase flows. The present study focuses on the phenomenological and mathematical modeling of the processes in the injector from the viewpoint of its non-equilibrium nature.
Date: December 31, 1993
Creator: Anand, G.
Partner: UNT Libraries Government Documents Department

Experimental investigation of sedimentation of LOCA - generated fibrous debris and sludge in BWR suppression pools

Description: Several tests were conducted in a 1:2.4 scale model of a Mark I suppression pool to investigate the behavior of fibrous insulation and sludge debris under LOCA conditions. NUKON{trademark} shreds, manually cut and tore up in a leaf shredder, and iron oxide particles were used to simulate fibrous and sludge debris, respectively. The suppression pool model included four downcomers fitted with pistons to simulate the steam-water oscillations during chugging expected during a LOCA. The study was conducted to provide debris settling velocity data for the models used in the BLOCKAGE computer code, developed to estimate the ECCS pump head loss due to clogging of the strainers with LOCA generated debris. The tests showed that the debris, both fibrous and particulate, remains fully mixed during chugging; they also showed that, during chugging, the fibrous debris underwent fragmentation into smaller sizes, including individual fibers. Measured concentrations showed that fibrous debris settled slower than the sludge, and that the settling behavior of each material is independent of the presence of the other material. Finally, these tests showed that the assumption of considering uniform debris concentration during strainer calculations is reasonable. The tests did not consider the effects of the operation of the ECCS on the transport of debris in the suppression pool.
Date: December 1, 1995
Creator: Souto, F.J. & Rao, D.V.
Partner: UNT Libraries Government Documents Department

Application of PSA to review and define technical specifications for advanced nuclear power plants

Description: As part of the design certification process, probabilistic safety assessments (PSAS) are performed at the design stage for each advanced nuclear power plant. Among other usages, these PSAs are important inputs in defining the Technical Specifications (TSs) for these plants. Knowledge gained from their use in improving the TSs for operating nuclear power plants is providing methods and insights for using PSAs at this early stage. Evaluating the safety or the risk significance of the TSs to be defined for an advanced plant encompasses diverse aspects: (a) determining the basic limiting condition for operation (LCO); (b) structuring conditions associated with the LCO; (c) defining completion times (equivalent to allowed outage times in the TS for conventional plants); and, (d) prescribing required actions to be taken within the specified completion times. In this paper, we consider the use of PSA in defining the TSs for an advanced nuclear plant, namely General Electric`s Advanced Boiling Water Reactor (ABWR). Similar approaches are being taken for ABB-CE`s System 80+ and Westinghouse`s AP-600. We discuss the general features of an advanced reactor`s TS, how PSA is being used in reviewing the TSs, and we give an example where the TS submittal was reviewed using a PSA-based analysis to arrive at the requirements for the plant.
Date: November 1, 1995
Creator: Kim, I.S.; Samanta, P.K.; Reinhart, F.M. & Wohl, M.L.
Partner: UNT Libraries Government Documents Department

Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina nuclear power plant.

Description: Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: GDH impact on an adjacent GDH and its attached piping; and GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Date: February 26, 2002
Creator: Dundulis, G.; Kulak, R.F.; Marchertas, A.; Narvydas, E.; Petri, M.C. & Uspusas, E.
Partner: UNT Libraries Government Documents Department

GDH pipe break transient analysis of the RBMK - 1500.

Description: Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. The cases of GDH impact on an adjacent GDH and its attached piping are investigated in this paper. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.
Date: May 15, 2002
Creator: Kulak , R.; Marcherta, A. & Dundulis, G.
Partner: UNT Libraries Government Documents Department

Intermediate-break LOCA analyses for the AP600 design

Description: A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has been analyzed. This event is characterized as an intermediate break loss-of-coolant accident (IBLOCA). Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PFl/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. The key processes occurring in an AP600 during a IBLOCA are primary coolant system depressurization, inventory depletion, inventory replacement via emergency core coolant injection, continuous core cooling, and long-term decay heat rejection to the atmosphere. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated Thus, the observation that the core is continuously cooled should be verified for the latter phase of the long-term cooling period, the interval when sump injection and containment cooling processes are important.
Date: July 1, 1995
Creator: Boyack, B.E. & Lime, J.F.
Partner: UNT Libraries Government Documents Department

Estimated net value and uncertainty for automating ECCS switchover at PWRs

Description: Question for resolution of Generic Safety Issue No. 24 is whether or not PWRs that currently rely on a manual system for ECCS switchover to recirculation should be required to install an automatic system. Risk estimates are obtained by reevaluating the contributions to core damage frequencies (CDFs) associated with failures of manual and semiautomatic switchover at a representative PWR. This study considers each separate instruction of the corresponding emergency operating procedures (EOPs), the mechanism for each control, and the relation of each control to its neighbors. Important contributions to CDF include human errors that result in completely coupled failure of both trains and failure to enter the required EOP. It is found that changeover to a semiautomatic system is not justified on the basis of cost-benefit analysis: going from a manual to a semiautomatic system reduces the CDF by 1.7 {times} 10{sup {minus}5} per reactor-year, but the probability that the net cost of the modification being less than $1, 000 per person-rem is about 20% without license renewal. Scoping analyses, using optimist assumptions, were performed for a changeover to a semiautomatic system with automatic actuation and to a fully automatic system; in these cases the probability of a net cost being less than $1,000/person-rem is about 50% without license renewal and over 95% with license renewal.
Date: February 1, 1996
Creator: Walsh, B.; Brideau, J.; Comes, L.; Darby, J.; Guttmann, H.; Sciacca, F. et al.
Partner: UNT Libraries Government Documents Department

SATURATED-SUBCOOLED STRATIFIED FLOW IN HORIZONTAL PIPES

Description: Advanced light water reactor systems are designed to use passive emergency core cooling systems with horizontal pipes that provide highly subcooled water from water storage tanks or passive heat exchangers to the reactor vessel core under accident conditions. Because passive systems are driven by density gradients, the horizontal pipes often do not flow full and thus have a free surface that is exposed to saturated steam and stratified flow is present.
Date: August 1, 2010
Creator: Schultz, Richard
Partner: UNT Libraries Government Documents Department

Final report, BWR drywell debris transport Phenomena Identification and Ranking Tables (PIRTs)

Description: The Nuclear Regulatory Commission has issued a Regulatory Bulletin and accompanying Regulatory Guide (1.82, Rev. 2) which requires licensees of boiling water reactors to develop a specific plan of action (including hardware backfits, if necessary) to preclude the possibility of early emergency core cooling system strainer blockage following a postulated loss-of-coolant-accident. The postulated mechanism for strainer blockage is destruction of piping insulation in the vicinity of the break and subsequent transport of fragmented insulation to the wetwell. In the absence of more definitive information, the Regulatory Guide recommends that licensees assume a drywell debris transport fraction of 1.0. Accordingly, the Nuclear Regulatory Commission initiated research focused toward developing a technical basis to provide insights useful to regulatory oversight of licensee submittals associated with resolution of the postulated strainer blockage issue. Part of this program was directed towards experimental and analytical research leading to a more realistic specification of the debris transport through the drywell to the wetwell. To help focus this development into a cost effective effort, a panel, with broad based knowledge and experience, was formed to address the relative importance of the various phenomena that can be expected in plant response to postulated accidents that may produce strainer blockage. The resulting phenomena identification and ranking tables reported herein were used to help guide research. The phenomena occurring in boiling water reactors drywells was the specific focus of the panel, although supporting experimental data and calculations of debris transport fractions were considered.
Date: September 1, 1997
Creator: Wilson, G.E.; Boyack, B.E.; Leonard, M.T.; Williams, K.A. & Wolf, L.T.
Partner: UNT Libraries Government Documents Department

Evaluation of emergency cooling water system

Description: Evaluation of the adequacy of the emergency cooling water addition system (ECWA system) requires analysis of the postulated accidents for which the system is required to function. Analysis of these accidents requires a knowledge of the amount of the ECW added that goes to the fuel; both the total amount to the fuel and the amount to the central fuel assemblies. This memorandum presents the methods used to calculate the flows to the fuel and the results of the calculations. The calculations are illustrated with a loss of pumping power accident and a plenum line break accident.
Date: October 1, 1968
Partner: UNT Libraries Government Documents Department

Improvements to the RELAP5/MOD3 reflood model and uncertainty quantification of reflood peak clad temperature

Description: Assessment of the original REAP/N4OD3.1 code against the FLECHT SEASET series of experiments has identified some weaknesses of the reflood model, such as the lack of a quenching temperature model, the shortcoming of the Chen transition boiling model, and the incorrect prediction of droplet size and interfacial heat transfer. Also, high temperature spikes during the reflood calculation resulted in high steam flow oscillation and liquid carryover. An effort had been made to improve the code with respect to the above weakness, and the necessary model for the wall heat transfer package and the numerical scheme had been modified. Some important FLECHT-SEASET experiments were assessed using the improved version and standard version. The result from the improved REAP/MOD3.1 shows the weaknesses of REAP/N4OD3.1 were much improved when compared to the standard MOD3.1 code. The prediction of void profile and cladding temperature agreed better with test data, especially for the gravity feed test. The scatter diagram of peak cladding temperatures (PCTs) is made from the comparison of all the calculated PCTs and the corresponding experimental values. The deviation between experimental and calculated PCTs were calculated for 2793 data points. The deviations are shown to be normally distributed, and used to quantify statistically the PCT uncertainty of the code. The upper limit of PCT uncertainty at 95% confidence level is evaluated to be about 99K.
Date: October 1, 1996
Creator: Chung, Bub Dong; Lee, Young Lee; Park, Chan Eok & Lee, Sang Yong
Partner: UNT Libraries Government Documents Department

Experiment DTA report for semiscale transparent vessel countercurrent flow tests

Description: Steady state air-water tests were performed as part of the Semiscale Blowdown and Emergency Core Cooling (ECC) Project to investigate downcomer countercurrent flow and downcomer bypass flow phenomena. These tests were performed in a plexiglass representation of the Semiscale pressure vessel which allowed changes to be madein the geometry of the upper annulus and downcomer for the purpose of investigating the sensitivity of downcomer and bypass flow to changes in system geometry. Tests were also performed to investigate the effects of two-phase inlet flows and different initial system pressures on countercurrent and bypass flow. Results for each test are presented in the form of computer printout of the measurements and of a summary of the pertinent calculated flow rates, pressures, and dimensionless volumetric fluxes. Descriptions of the test facility, instrumentation, operating procedures, and test conditions are also presented. An error analysis is presented for selected volumetric flux calculations. 10 references. (auth)
Date: October 1, 1975
Creator: Hanson, D.J.
Partner: UNT Libraries Government Documents Department