Tests in the ATLE Loop on the PIUS design
Description
The self-protective thermal hydraulic design of the primary system that prevents reactor power from exceeding the coolant capability of the submerging water independent of outside surveillance and supervision is a fundamental feature of the PIUS design. The experimental demonstration of the self-protective features of the primary system in a large scale test loop (the ATLE Loop) in the ABB Atom Engineering Laboratories are described in the paper. The volume scale of the ATLE Loop is approximately 1:300. The height of the test loop is the same as in the full scale design to ensure similarity in the driving buoyancy force during a transient. The test loop includes the following: one electrically heated, simulated fuel assembly, reactor pool, primary systems, upper and lower density locks and heat exchangers. System response to various severe transients and normal operations was studied. Comparisons were made with the predictions of the RIGEL code, which has been developed by ABB Atom specifically for study of the PIUS design and its predecessor designs. Selected results of these studies and comparison of these results with predictions obtained using RIGEL are presented in this paper
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Additional details
Publishing Information
- Imprint Title
- Transactions of the nineteenth water reactor safety information meeting
- Imprint Pagination
- 220 p.
- Journal Page Range
- p. 3.1.
- Report number
- NUREG/CP--0118
Conference
- Title
- 19. Nuclear Regulatory Commission (NRC) water reactor safety information meeting.
- Dates
- 28-30 Oct 1991.
- Place
- Bethesda, MD (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 23059427
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BORON; BWR TYPE REACTORS; COMPUTER CODES; COMPUTERIZED SIMULATION; DEMONSTRATION PROGRAMS; DESIGN; FLOW RATE; HEAT TRANSFER; HYDRAULICS; MONITORING; PRESSURE DEPENDENCE; PRIMARY COOLANT CIRCUITS; PUMPS; PWR TYPE REACTORS; R CODES; REACTOR PROTECTION SYSTEMS; REACTOR SAFETY; REACTOR SAFETY EXPERIMENTS; SCALE MODELS; STANDARDIZATION; TEMPERATURE DEPENDENCE; TEST FACILITIES; TRANSIENTS; VERIFICATION
- Descriptors DEC
- COOLING SYSTEMS; ELEMENTS; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; EQUIPMENT; POWER REACTORS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; SAFETY; SEMIMETALS; SIMULATION; STRUCTURAL MODELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-911079--.