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PFR evaporator leak

Abstract

PFR has three heat removal circuits each one having an evaporator, superheater, reheater; all separate units. The status of the system was that circuit No 3 was steaming with 10 MW thermal nuclear power; No 1 circuit was filled with sodium but with the evaporator awaiting modification to cure gas entrainment problems already reported. The leak was in No 2 circuit and was located in the evaporator unit. The evaporator is rated at 120 M thermal at full power and as such is a large unit. The circuit was filled with both sodium and water for the first time three weeks before the conference so it was recent history being reported and therefore any figures quoted should be taken as indicative only. The history of the steam generator was that it was built at works to a very high standard and underwent all the usual tests of strength, inspection of welds and helium leak testing. The steam generator is of U tube design with a tube plate to which the boiler tubes are welded, with all the welds in one of two gas spaces. The inlet and outlet sides are separated by a baffle and the salient features are illustrated  More>>
Authors:
Publication Date:
Jul 01, 1975
Product Type:
Conference
Report Number:
IWGFR-1
Reference Number:
EDB-00:099587
Resource Relation:
Conference: Study group meeting on steam generators for LMFBR's, Bensberg (Germany), 14-17 Oct 1974; Other Information: PBD: Jul 1975; Related Information: In: Study group meeting on steam generators for LMFBR's. Summary report, 334 pages.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; AFTER-HEAT REMOVAL; EVAPORATORS; LEAK TESTING; PFR REACTOR; SODIUM; STEAM GENERATORS; SUPERHEATERS; WELDING
OSTI ID:
20107705
Research Organizations:
International Atomic Energy Agency, International Working Group on Fast Reactors, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
TRN: XA0055725054542
Availability:
Available from INIS in electronic form
Submitting Site:
INIS
Size:
page(s) 299-302
Announcement Date:
Dec 01, 2000

Citation Formats

Smedley, J A. PFR evaporator leak. IAEA: N. p., 1975. Web.
Smedley, J A. PFR evaporator leak. IAEA.
Smedley, J A. 1975. "PFR evaporator leak." IAEA.
@misc{etde_20107705,
title = {PFR evaporator leak}
author = {Smedley, J A}
abstractNote = {PFR has three heat removal circuits each one having an evaporator, superheater, reheater; all separate units. The status of the system was that circuit No 3 was steaming with 10 MW thermal nuclear power; No 1 circuit was filled with sodium but with the evaporator awaiting modification to cure gas entrainment problems already reported. The leak was in No 2 circuit and was located in the evaporator unit. The evaporator is rated at 120 M thermal at full power and as such is a large unit. The circuit was filled with both sodium and water for the first time three weeks before the conference so it was recent history being reported and therefore any figures quoted should be taken as indicative only. The history of the steam generator was that it was built at works to a very high standard and underwent all the usual tests of strength, inspection of welds and helium leak testing. The steam generator is of U tube design with a tube plate to which the boiler tubes are welded, with all the welds in one of two gas spaces. The inlet and outlet sides are separated by a baffle and the salient features are illustrated in the attached figure. The unit achieved a leak tightness better than the detection limit in the helium leak test at works. This limit was assessed as being less than an equivalent leak of 10{sup -6} g/s water under steam generator service conditions. However even though all the steam generator units passed this test at works a further test was carried out when the circuits had been completed. The test was carried out during commissioning after sodium filling and with the units hot. The method was to introduce a mixture of helium/ argon at 500 pounds/square inch into the water side of the steam generators and measure the helium concentration in the sodium side gas spaces of the circuit. The test lasted many days and under these conditions the sensitivity is such that a leak equivalent to somewhere between 10{sup -7} to 10{sup -6} g/s equivalent water leak could be detected, i.e. well down into the micro leak region.}
place = {IAEA}
year = {1975}
month = {Jul}
}