Published 2000 | Version v1
Miscellaneous

Numerical simulation of the RD-14M test T9308

  • 1. Atlantic Nuclear Services Ltd., New Brunswick (Canada)
  • 2. NB Power, New Brunswick (Canada)

Description

In 1990, the AECB opened the Action Item 90G02 to track the progress of the Canadian nuclear utilities in addressing concerns related to core cooling in the absence of forced flow. The main concern is that, under thermosyphoning conditions in the unbroken loop following a LOCA, flow may stop in individual channels. This may lead to high sheath temperatures, possibly jeopardising the integrity of the unbroken loop pressure boundary and an escalation of the initial event consequences. This type of flow behaviour occurred in a few thermosyphoning experiments performed with the RD-14M facility [1] at relatively high system inventory. The percentage of inventory in the loop at this time was comparable to that in the reactor unbroken loop following a LOCA. Over the intervening years, many simulations of the RD-14M experiments in which thermosyphoning flow breaks down have been performed. Recently, simulations of one test, T9308 [2] have shown that individual heated section behaviour is best matched with various two-phase conditions at each feeder connection to the outlet header. However, the header conditions in these CATHENA simulations were treated as uniform (single node boundary condition) and trial and error was used to determine the transient void conditions. A more fundamentally based assessment of the header fluid conditions is therefore required in order to predict the reactor situation confidently. The purpose of this work is to simulate the transient 3-D behaviour of the water-steam mixture within the outlet header no.5 of the RD-14M loop during Test T9308. The header geometry is shown in Figure 1, in which five outlet feeders from heated sections 10 to 14 are connected to the header. Also attached to the header are one boiler pipe rising to boiler no.1, and one dead-ended pipe to a pressure relief valve. This pressure relief valve is not representative of a CANDU, but is required for over-pressure protection of the RD-14M test-rig. The dead-ended pipe has been modified to save computer resources while conserving the volume and height of the original pipe, and it may be important during the transient process for refilling the header due to its contained volume and height relative to the header. (author)

Part of:
A better nuclear tomorrow: 21. annual CNS conference

Additional details

Publishing Information

Publisher
Canadian Nuclear Society
Imprint Place
Toronto, Ontario (Canada)
ISBN
0-919784-66-6
Imprint Title
A better nuclear tomorrow: 21. annual CNS conference
Imprint Pagination
49.1 Megabytes
Journal Page Range
[3 p.]

Conference

Title
Proceedings of the Canadian Nuclear Society 21st annual conference
Dates
11-14 Jun 2000
Place
Toronto, Ontario (Canada)

INIS

Country of Publication
Canada
Country of Input or Organization
Canada
INIS RN
35088031
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference, Non-conventional Literature
Descriptors DEI
CANADA; CANADIAN AECB; COMPUTERIZED SIMULATION; FUEL CANS; LOSS OF COOLANT; NUCLEAR FACILITIES; REACTOR CHANNELS; TWO-PHASE FLOW
Descriptors DEC
ACCIDENTS; CANADIAN ORGANIZATIONS; DEVELOPED COUNTRIES; FLUID FLOW; NATIONAL ORGANIZATIONS; NORTH AMERICA; REACTOR ACCIDENTS; REACTOR COMPONENTS; SIMULATION

Optional Information

Notes
3 refs., 1 fig.