Failure Knowledge Database / 100 Selected Cases

Leakage of primary coolant at Mihama Unit 2
due to failure of SG tube
February 9th, 1991, Mihama in Fukui Prefecture
KITSUNAI, Yoshio (Japan Crane Association)
KOBAYASHI, Hideo (Tokyo Institute of Technology)(Summary)
On February 9th, 1991, a heat transfer tube (SG tube) in a steam generator of the No.2 pressurized water
reactor a t the Mi hama nuc lear power st ation of t he Kansai El ectric Power C ompany br oke of f during a
rated output operation. As a result, about 55 tons of primary cooling water leaked out from the SG tube into
the s econdary cooli ng loop, and t he re actor was scra mmed by op eration of the E CCS (Em ergency CoreCooling System). The failure of the SG tube was caused by fretting fatigue resulting from contact of the SG
tube with the supporting plate for the SG tubes, because the AVB, which functions to prevent flow-induced
vibration, was not inserted deep enough onto the SG tubes in the steam generator. The scale of the accident
was ranked "level 3" on the international nuclear events scale (INES).
1. Component
Steam generator of pressurized water reactor
Figure 1 s hows a sche matic i llustration of t he pressurized wat er reactor. Figure 2 s hows the ste amgenerator and the location of the failure of the SG tube.
The s team generat or is a pressure vessel with a heat exchanger. T he st eam generator has an outer
diameter of arou nd 4 m and a hei ght of ar ound 20m. 3260 SG tubes were installed in t he heat exchanger.
The tubes, each with an outer diameter of 22.2 mm and a thickness of 1.27 mm, were made of Inconel 660
material. Most of the l ength of the SG t ubes was stra ight, but the upper p art of th e t ubes i n the he atexchanger was bent i n an opposite U-shape with certain curvatures. The straight part of t he SG tubes washeld at six positions by supporting plates. At the s ixth supporting posi tion near the up per part of t he SG
tubes, a specifically shaped (V-shaped) anti-vibration bar, AVB, was installed on the bent part of SG tubes
to prevent flow-induced vibration.
2. Event
At around 13: 50, on February 9th, 1991, leakage of about 55 tons of primary coolant occurred due tothe failure of a SG tube i n a steam generator of the No.2 pressurized water reac tor at t he Mihama nuclear
power sta tion of the Kans ai El ectric Power Com pany. The r eactor shut down c ommenced im mediately
triggered b y a si gnal warni ng th at t he primary c oolant i n t he reactor was d ecreasing. The ESSC wasoperated im mediately, a nd about 51 tons of co olant w ater was flo oded into the r eactor. The amount ofsteam released fro m the m ain steam relief v alve t o atm osphere was abou t 1. 3 to ns. The amounts ofradioactive ra re gas a nd io dine discharged to t he atmosphere w ere ab out 2.3E10 a nd 3 .4E8 becquerels,
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Failure Knowledge Database / 100 Selected Cases

respectively.
After the accident, investigation of the st eam generator w as carried out using a fiber sco pe and so me
other inspection instruments. As a result, the failure of a SG tube was found near the sixth supporting plate
for the SG tubes.
3. Course
At 13:40, an alarm of a condenser air off take system went off during a rated output operation, warningthat t he c oolant w ater level in t he s team ge nerator w as de creasing. At 13 :50, an automatic em ergency
shutdown of t he reactor was triggered by the signal of decreasing pressure in the pressurizer. After sevenseconds, the ECCS was automatically operated, and coolant water was flooded into the reactor by a h ighpressure injection pump. However, one main steam isolation valve and on e pressurizer relief valve coul d
not be operated by remote control. Therefore, the valve operation was carried out manually.
4. Cause(1) Fracture surface
The f ailed t ube was rem oved from the heat e xchanger, a nd t he fr acture s urface was e xamined by ascanning electron m icroscope. S triations, w hich are a characteristic of fati gue fai lure, w ere observed o nlarge portions of th e fracture surface, and dimples showing tensile fracture were also observed. However,few traces of stress corrosion cracking and corrosion were found on the fracture surface of the tube. The
failure of the tube was, therefore, hypothesized to be due to cyclic loading. Figure 3 shows the morphologyof the fra cture surface of t he SG tube. F igure 4 shows a typic al e xample of the st riations formed on thefracture surface of the SG tube. Examination of the other SG tubes near the failed tube showed traces ofwearing f ormed by frett ing due to contact b etween t he tu bes a nd the a nti-vibration bars on the ou ter
surfaces of the tubes. Stress amplitude of the failed tube estimated based on the striation spacing was foundto be in the range of around 51 to 60 MPa.
(2) Fatigue fracture
Occurrence of cy clic loading i n the SG t ube t hat h ad faile d w as rela ted t o t he i nsertion depth o f t heanti-vibration bar, AVB. The SG tubes w ere subjected to vibrations due to the flow of seco ndary coolant
outside of the SG tubes. In order to prevent the flow-induced vibration, V-shaped AVBs were installed onto
the opposite U-bent SG tubes near the upper part of the steam generator. However, the insertion depth of
the AVB for the SG tu bes was not e nough, because the e ngineers w ho ins talled the A VB di d not f ully
understand the importance of the AVB. In fac t, no damage w as founding in the SG tubes into w hich theAVB w ere in serted to suf ficient depth as show n by t he des ign guidelines. Fi gure 5 s hows t he insertion
position of the A VB on t he SG tub es by soli d li nes. A ccordingly, th e SG tu bes w ere su bjected toflow-induced vibration and strongly contacted with the sixth supporting plate, so that the SG tubes incurred
damage by fretti ng fati gue. Inspec tion of the AVB had no t b een carried out sin ce inst allation. Figure 6
shows a schematic illustration of the fatigue failure of the SG tube due to flow-induced vibration.

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Failure Knowledge Database / 100 Selected Cases

5. Immediate Action
After t he ac cident, a detailed ex amination o f th e AVB fo r th e SG tu bes in th e st eam g enerators of a llreactors in t he Kansai El ectric Company was carried out. As a resul t, lack of suf ficient insertion depth of
the AVB was found in some heat exchangers. These AVBs were replaced with new ones and installed at thedesignated d epth in t he st eam genera tors. More over, t he st eam generator of t he No.2 pressurized w ater
reactor was also replaced with a new one, because many of the SG tubes were removed from the generator
for failure analysis.
6. Countermeasure(1) Inspection of the installation of the AVB for t he SG tubes and of the mounting position of the AVB
before operation.
(2) Execution of regular inspections of the AVB.
(3) Inspection for distortion and damage of the supporting plate for the SG tubes.
(4) Development of a new detecting system that can quickly and accurately detected signs of damages of
the SG tubes in a steam generator.
(5) Development of a new type of AVB with high performance and easy installation.
(6) Employers should make sure th at the engineers who are engaged in fabrication or maintenance of thedevices and equipment in nuclear reactors understand the function of those devices and equipment.
7. Knowledge
A discrim inate eq uation describing t he ge neration of flow -induced vibrations for SG tu bes w assuggested based on a maintenance standard of the Japan Society of Mechanical Engineers, as follows.
SR = Ue/Uc < 1
Where SR is the discriminate value, Ue is the effective velocity, and Uc is the critical velocity.
8. On the Side
In order to provide an opportunity to learn from the accident that resulted in th e leakage of primarycoolant from the SG tu be due t o frett ing fatigue, the d amaged ste am gen erator has bee n preser ved i n a nexhibition at the Mihama station of the Kansai Electric Power Company. An exhi bition is a good way tohelp everyone to good lessons from an accident.
9. Social Impact
This a ccident was the f irst disaster i n Ja pan that re sulted in actuation of the EC CS due t o le akage ofprimary coolant in the steam generator. Therefore, the accident caused social concern with nuclear reactors.
The i nternational nuclear ev ents sc ale (IN ES) is define d by th e IA EA to assure c oherent re porting ofnuclear accident by different official authorities. The INES is chara cterized from level o ne to level seven.
The l evel number increase s with t he s cale of the ac cident. For example, le vel one is a m inor eve nt, an dlevel seven is major accident. The scale of the accident in 1979 resulting in the loss of coolant that occurredin Three Mile Island was ranked level five by the IAEA. The accident reported here was ranked level three.
3Failure Knowledge Database / 100 Selected Cases

10. Information Source(1) Nuclear Power Engineering Test Center, Nuclear Power Safety Infor mation Research Center, On thedamage of SG tube in steam generator of No.2 pressurized water reactor at Mihama Station of Kansai
Electric Power Company, (1992)(2) htt p://www.atom.meti.go.jp/atom-db/jp/index.html(3) Maintenance standard of the Japan Society of Mechanical Engineers, JSME S016-2002
11. Primary Scenario
01. Ignorance
02. Insufficient Knowledge
03. Poor Understanding
04. Production
05. Hardware Production
06. Production of Machinery and Equipment
07. Heat Exchanger
08. SG tube
09. Supporting Plate
10. AVB
11. Installation
12. Regular Operation13. Nonobservance of Procedure
14. Error of Mounting Position
15. Lack of Insert Depth
16. Failure
17. Fracture/Damage
18. Flow-induced Vibration
19. Fretting Fatigue
20. Failure of SG tube
21. Usage
22. Maintenance/Repair
23. Inspection
23. Lack of Inspection
24. Failure
25. Large-Scale Damage
26. Leakage of Coolant

4Failure Knowledge Database / 100 Selected Cases

Fig. 1

Schematic illustration of pressurized water reactor.

Fig. 2

Steam generator and location of failure of SG tube.

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Failure Knowledge Database / 100 Selected Cases

Fig. 3Fig. 4Morphology of fracture surface of the SG tube.

Striation formed on the fracture surface of the SG tube.

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Failure Knowledge Database / 100 Selected Cases

Solid lines show the location of AVB when accident occurred.
Dotted lines show normal position of the AVB.
Fig. 5

Fig. 6

Location of AVB on the SG tubes.

Schematic illustration of fatigue failure of the SG tube due to flow-induced vibration.

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