ML19317D714
| ML19317D714 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/04/1972 |
| From: | Thies A DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19317D712 | List: |
| References | |
| NUDOCS 7912100488 | |
| Download: ML19317D714 (7) | |
Text
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DURE P0krER COMPAW OCONEE NUCLEAR STATION Regulatory File Cy.
UNIT 1
- r. a?:ss w/ttr ten " ~ 'T - 7" REACTOR COOLANT SYSTEM INCIDENT REPORT l
0 DOCKET No. 50-269 APRIL 4, 1972
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TABLE OF CONTENTS I.
Summary II.
Sequence of Events III.
Extent of the Damage IV.
Status of Recovery Operation LIST OF FIGURES Fisture No.
1.
Reactor Coolant System Pressure / Temperature History (2/21/72 to 3/10/72) 2.
Reactor Vessel and Internals -
General Arrangement i
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DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT NUMBER 1 REACTOR C00LAhT SYSTEM INCIDENT REPORT I.
Summarv At the conclusion of the first phase of the hot functional testing program of Oconee Station Unit 1 on March 11, 1972, an inspection of the reactor coolant system revealed the failure of several reactor internal components.
The loose parts resulting from these failures caused extensive damage to the tube ends and to the tube sheet welds in the "A" steam generator upper head.
The "B" steam generator upper head tube ends and welds experienced only minor damage.
It has been determined that these parts are principally the 3/4" diameter incore instrument nozzles which penetrate the bottom of the reactor vessel and allow insertion of flux instrumentation into the reactor core.
Twenty-one incore instrument nozzles broke off just above the weld on the inner surface of the reactor vessel. The cause of these failures and the schedule for repair has not yet been determined. This report presents a summary of the pertinent events related to these failures and describes the Jamage observed in the reactor coolant system during the subsequent inspection.
II.
Secuence of Events This section of the report presenis a chronological sequence of events which are pertinent to the observed reactor coolant system damage, as derived from shif tlogs and discussions with test and operating personnel.
In addition, a plot, Figure 1, is presented of the temperature and pressure during the first phase of the hot functional test from February 21 through March 10, 1972.
1.
On March 3, 1972, a noise was heard on the "A" loop side of the reactor
- I'F coolant system at a temperature of appro-imately 350*F.
No noise was CID'.
heard on the "B" loop side. The noise a;opped when both "A" loop reactor coolant pumps were secured. Also on March 3, 1972, the signal j {"^
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from accelerometer SN104 located in the thermocouple guide tube began to show 60 Hz noise, indicating an open lead.
B&W project management dko I' personnel were notified and arrangements were made for B&W engineering "[b personnel to visit the station the next day.
2.
On March 4, 1972, heat up from 500*F to 532*F was started.
The noise level was approximately the same as observed on March 3.
Of 60 tem-porary thermocouples installed in the once-through steam generator "B"
(OTSG "B"), two now showed a slight divergence reading. Representatives of B&W engineering arrived at the site to monitor the noise levels in OTSG "A".
Opinions varied among the observers as to whether the noise was louder at the top or bottom of the steam generator, but were con-sistent in locating the source of the noise within the OTSG "A".
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3.
On March 5, 1972, the noise level in OTSG "A" seemed lower.
4.
On March 6 engineering personnel continued to assess the source and character of the noise, and arrangements were made to bring noise analysis equipment to the site.
Also on March 5 and 6 reactor coolant system leakage measurements were being made; the results of these calculations indicated no uausual leakage.
i 5.
Investigations on March 7, 1972, revealed that 23 of the 60 temporary thermocouples in OTSG "B" had failed.
Personnel from B&W Alliance Research Center arrived to record and analyze the noise for various reactor coolant pump combinations.
6.
On March 8, 9, and 10, the noise levels in OTSG "A" were essentially the same as observed on March 5 and 6.
No specific conclusion could be reached regarding the source of the noise. OTSG "B" was checked frequently from March 3 through 10 and no noise was heard.
7.
On March 10, 1972, the first phase of het functional testing had been completed and the reactor coolant system was cooled down to 250'F and pressure reduced approximately 400 psig.
The reactor coolant system was depressurized and drained on March 11 and 12.
On March 13, 1972, the upper manway on OTSG "A" was opened and metallic objects of various shapes were seen on the tube sheet. The upper manway on OTSG "B" was also opened and inpsection revealed approximately one-half of the temporary thermocouple conduit (which is part of the first-of-a-kind instrumentation for OTSG testing) was dislocated and four pieces of metallic tubular material were seen on top of the tube sheet.
8.
On March 14, 1972, the lower manways on both steam generators were opened; small flakes of metallic material were noted on the lower head surfaces.
The decision was then made to ret 7e the reactor vessel head.
9.
On March 17, 1972, the reactor vessel head was removed. Metal flakes were noted on the top of the plenum assembly. Two 1/2" diameter thermocouple guide tubes were observed to be missing from the upper plenum assembly (one contained accelerometer SN104 and the other was empty).
The upper plenum and internals were subsequently removed on March 17 and 18 and a number of incore instrument nozzles and guide tubes were noted to be broken.
III.
Extent of the Damage Inspection of the reactor coolant system revealed the following damage as of March 30, 1972:
(Figure 2 illustrates the location of the damage for numbers 3, 4, 5, 7, 8, and 9 below) 1.
In OTSG "A" extensive impact damage to the tube sheet welds was observed on the upper tube sheet.
This damage apparently was caused by repeated impact of the metallic objects (principally pieces of the.
r' 3/4" diameter incore instrument nozzles) which were found in the upper plenum of the steam generator. The inside surface of the steam generator head also showed the effects of impact.
2.
In OTSG "B" about half of the tube ends showed indication of being hit.
Approximately 10 percent of these will require minor veld repair.
The metal objects found in the upper plenum of this steam generator were trapped in the temporary instrumentation, apparently the reason for the limited extent of damage to the tube ends and tube to tube sheet welds.
There are some impact marks on the inside of the OTSG "B".
The temporary thermocouple instrumentation was partially destroyed.
3.
Twenty-one 3/4" diameter incore instrument nozzles were broken off.
Fourteen additional nozzles were cracked in the region above the weld.
Seventeen nozzles showed no cracks.
4.
Fear incore instrument guide tube extensions were broken off below the lower flow distributor and their four spider assemblies were dis-located.
One of these was also broken above the distributor. Four additional guide tubes were cracked in the region of the weld below the flow distributor. Forty-four tubes showed no cracks.
5.
There are nicks and scratches on the lower flow distributor.
6.
Nicks and impact marks were observed on the flow guide vanes and the bottom of the reactor vessel.
7.
At the mating surfaces between the thermal shield and the lower grid assembly, some metal upset at the edges was observed.
The retention weld on all eight dowels at the lower edge of the thermal shield were broken and one of the dowels had backed out approximately 3/4".
8.
Two of six installed 1/2" diameter thermocouple guide tubes were broken off of the upper plenum assemtly.
9.
There are minor scratches on the face seal between the reactor vessel 36" outlet nozzle and the core support shield.
10.
Minor scratch marks were noted in both the 36" and the 28" reactor coolant piping.
11.
Minor scratches were observed on the reactor coolant pump impellers.
IV.
Status of Recovery Operations Repairs are underway on the OTSG "B" using approved factory repair pro-cedures while on OTSG "A" investigation of damage and development of repair procedures continues.
An orderly investigation will determine the cause of the failures in the reactor vessel and will establish the basis for appropriate re-design and repairs. Approved procedures will be used during the investigations and repair work.. _ - - _ _
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4 REACTOR VESSEL AND INTERNALS -
GENERAL ARRANGEMENT OCONEE NUCLEAR STATION Figure 2 1