Summary of the Semiscale Program, 1965-1986

Loomis, G.G.

doi:10.2172/6302855

Cited by 16 publications

(9 citation statements)

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“…Steam generator (SG) secondary-side depressurization through the SG valve(s) with auxiliary feedwater (AFW) injection is one of major accident management (AM) measures to cool and depressurize the primary system because of steam condensation in the SG U-tubes to introduce lowpressure injection (LPI) system of ECCS during SBLOCA especially when HPI system is totally failed [6], as shown 2 International Journal of Nuclear Energy in Figure 1. There have been some experimental studies of SBLOCAs with AM measures by such integral test facilities as Semiscale [7], LOBI [8] and PKL [9]. The obtained data, however, were not sufficient to clarify the effectiveness of AM measures because of such atypical features as small volume and low pressure in the primary system.…”

Section: Introductionmentioning

confidence: 97%

RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

Takeda

2014

International Journal of Nuclear Energy

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RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM) measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW) injection into the secondary-side of both steam generators (SGs) as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI) system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

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Section: Introductionmentioning

confidence: 97%

RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

Takeda

2014

International Journal of Nuclear Energy

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“…This was probably not a good design approach from a code validation standpoint. After years of more careful consideration and development of the electrically-heated Semiscale reactor simulator (Loomis, 1987) beginning in the late 1960s, the emphasis for LOFT was shifted to the investigation of loss-of-coolant accidents (LOCAs) and emergency core coolant system performance. Concurrently, computer models that predicted the fluid behavior of LOCAs, such as RELAP, were developed at the site.…”

Section: Iii1 Historical Capabilitiesmentioning

confidence: 99%

“…The buildings are owned by CWI. Following the conclusion of the Semiscale program (Loomis, 1987) a USNRC commissioned study was conducted at the INL to define scaling concepts upon which future integral thermal hydraulic experimental facilities should be based (Condie, et al, 1987). Four separate scaling concepts were evaluated based on their ability to reproduce important PWR reactor thermal hydraulic phenomena.…”

Section: A33 Possible Integral Reactor Simulation Facilitiesmentioning

confidence: 99%

“…Experiments that model integral effects are typically carried out in large integral facilities, such as Semiscale (Loomis, 1987) and LOFT (Nalezny, 1983), that model the complete reactor primary system plus connected heat transfer and safety systems. Experiments that model phenomena in a component are often geometrically scaled to the prototype component and simulate the often three-dimensional "separate effects" within the component.…”

Section: Introductionmentioning

confidence: 99%

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INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

McCreery¹,

McIlroy²

2007

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“…This is the subject of major research support described in Section 5. '_'_l,,,1....... I.lJl,,ld,I,.. _,1 .................... ,,,11,_11 ...... _ ...................... ,,,Jllu,,mild.,l_ ................…”

Section: Cmentioning

confidence: 99%