Nuclear Safety Research

5-4Reactor Instruments to Prevent PWR Severe Accident

Depressurization byrelief valve (RV)
Pressurizer

Level
L meter

Normalized
Superheat (-)–LSTF Test to Estimate Usefulness of Core Exit Thermocouple (CET)–

0.4(a) Superheat of CET

0.2

1074 s,Start ofAM action

910 s,Start ofsuper-heat

0700

900

Min.
1000

1100

1.8

Normalized Rod
Temperature (-)LT(b) Rod surface temperature
at several elevations

1.4Gas

1.2

840 s,Start ofheat-up0.8700

Low pressure injection(LPI) (P<1.2MPa)
High pressure injection(HPI)

Fig.5-7 CET validation in a PWR top break LOCA test
LSTF is the largest PWR simulator in the world, with fullheight and 1/48 volume scaling. A top break test wasperformed, motivated by an incident of significant wall thinningaround the control rod penetration nozzle at pressure vessel
head of Davis Besse reactor (see Topic 5-1). Capability ofthe CET (“T” in Fig. 5-7) to detect core overheating in caseof an HPI failure was examined. The test revealed that theCET capability was influenced by three-dimensional steamflow from the core toward the CRGT inlets as well as theirlocation around the CRGT.

1500

1600

1215 s,Power limited
to 50%

P.5(Middle)

800

Tsat

P.3

900

1000

1100

1200

1300

1400

1500

1600

Time after Break (s)Fig.5-8 Maximum steam superheat detected by CET (a)was later and lower than superheat of rods in core (b)4.0Normalized Core Height (-)

Accumulator injection
system (AIS) (P<4.5MPa)

1400

P.7
P.9(Top)

1.0

L

1300

1.6

CRGT

PWR
vessel1200

Time after Break (s)Steamgenerator(SG)Vessel topbreak

800

Max.
Ave.

3.5
3.02.5(Top)P.9
t=1000 s
P.8
P.7
t=1100 s

P.6

2.0
1.5
1.0
0.5

t=1200 s

P.5(Middle)
P.4

Rod bundle with CRGT (B20)
Rod bundle w/o CRGT (B15)

P.3
Locationof T/Cs
00.1

0.2

0.3

0.4

0.5

0.6

0.7

Normalized Superheat (-)

Fig.5-9 Comparison of axial steam temperatures in two
rod bundles with and without CRGT at the core exit
Temperature profiles significantly changed with CRGT.

We conducted a Large Scale Test Facility (LSTF) test
simulating a pressurized water reactor (PWR) loss-of-coolant
accident (LOCA) caused by a small break at the vessel top
under assumption of the total failure of high pressure injection(HPI) system (Fig.5-7). This test was part of the OECD/NEA
ROSA Project which is an international cooperative study
on reactor safety and advanced code development, one goal
being the validation of a core exit thermocouple (CET) todetect core overheating and give a cue to start steam generator
depressurization, an accident management (AM) action.
The CET detected superheated steam temperature, but thiswas 70 s later than the start of core superheating, and there
was a further delay of 160 s to detect the temperature atwhich AM action is started, 623K. Thus, a large temperature
difference was found between the CET and the hottest fuel rod(Fig.5-8). We found two reasons for this. One was the cooling
effect of structural metals on uprising steam. The second one,specific for this test, was that steam flow convergence into thecontrol rod guide tube (CRGT) inlet (Fig.5-7) induced threedimensional steam flow in the core. In fact, steam temperatures
in upper part of a high power bundle without a CRGT (B15)
were significantly lower than those in B20 bundle with CRGT,
suggesting colder steam inflow from the periphery following
after the hot steam that went to a neighboring bundle with a
CRGT (Fig.5-9). The CET installation outside the CRGTs may
have caused further discrepancy from the hot core.
We showed that such CET response was possible in actual
PWR operation, and furthermore demonstrated the usefulnessof alternative indexes measured by water level meters (3 kinds;“L” in Fig.5-7).
OECD/NEA has just started study of CET effectiveness
in PWRs. (This paper contains findings of the OECD/ROSA
Project with its consent for publication.)

Reference
Suzuki, M. et al., Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experimentat OECD/NEA ROSA Project, Proceedings of 16th International Conference on Nuclear Engineering (ICONE16), Orlando, Florida, USA, 2008,
ICONE16-48754, 11p., in CD-ROM.

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JAEA R&D Review 2008