ML20050C845
| ML20050C845 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde, Humboldt Bay, 05000000 |
| Issue date: | 03/31/1982 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| Shared Package | |
| ML20050C837 | List: |
| References | |
| NUDOCS 8204090403 | |
| Download: ML20050C845 (1) | |
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s March 31, 1982 Docket tios. 50-528, 50-529, 50-530 Arizona Public Service Connany P. O. Box 21666 s
Phoenix, Arizona 85036 Attention!LMr.',E.E.VanBrunt,.1r.
<,.Vic<!: President, Nuclear Projects IEINFORMATIONf0TICE10.82-09: CRACKING IN PIPING OF MAKEUP COOLANT
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Gentlemen:
't The enclosed Informtf 0n Notice provides early notification of events that may have safety shnificance.
It is expected that recipients will review the Information Notice fcv possible applicability to their facilities.
Sincerely, OrYrdh-tnc, t
t R. H. Ergehen
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P.. H. Engelken Regional Administrator
Enclosure:
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8202040131 IN 82-09 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON,-D.C. 20555 l
March 31, 1982 IE INFORMATION NOTICE NO. 82-09: CRACKING IN PIPING 0F MAKEUP COOLANT LINES AT B&W PLANTS Description of Circumstances:
On - January 21, 1982, Crystal River Unit 3 commenced shutdown to investigate an unidentified 0.9-gpm primary leak.
During power reduction the leak rate increased to about 1.0 gpm and the plant proceeded to hot standby conditions.
A visual inspection inside the reactor building at this time revealed the leak was associated with a 2b-inch check valve (MOV-43) in the makeup line to the 26-inch reactor coolant (RC) loop A inlet line. This line is used for normal makeup of reactor coolant but is also part of the redundant high-pressure infec-tion system. After the insulation was removed from the affected valve-a 140
'r circumferential crack in the check valve body near the valve-to-safe end weld (i.e., valve end toward RC inlet nozzle) was found. The leak was nonisolatable and the plant promptly proceeded to cold shutdown conditions in accordance with plant technical specifications.
The check valve was removed and liquid penetrant testing (LPT) was performed oa the accessible inside diameter (ID) surfaces including ~5 inches into the 2h-inch line on the inlet side of the affected valve. This inspection dis-closed'an' extensive network of heat-check type cracks around the safe end ID surface. A similar condition was observed inside the valve body from the -
discharge si e up to the disc seat area. The valve inlet side and connecting piping were now affected. The most severe cracking in the safe end appeared to j
have penetrated up to 25 percent of the wall thickness. A visual inspection also revealed the thermal sleeve inside the high-pressure injection (HPI) i nozzle was loose and showed evidence of wear in areas of contact. Some cracking of the thermal sleeve was also observed.
l As a result of the Crystal River 3 findings, Duke Power Company initiated a l
radiographic examination of the RC inlet nozzle connections on the two HPI l
lines used for normal makeup at Oconee Unit 3 to determine the thermal sleeve conditions. This examination disclosed that in one of the makeup nozzles the thermal sleeve was loose, the four thennal sleeve retaining' button welds on the safe end side were missing, and the thermal sleeve was slightly displaced -in the upstream direction of flow. Action was~then taken to remove the pipe l
extension to replace the affected thermal sleeve.
Further findings and i
expanded inspection as a result of this action are suninarized below.
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i IN 82-09 fiarch 31, 1982 Page 2 of 3 Investigation and Findings:
A.
Crystal River A metallurgical investigation of the affected valve body indicated two crack initiation sites.
One was inside on the valve body at a machine mark (i.e., weld counterbore area) and one was on the outside diameter (0D) at the valve-to-weld transition (geometrical discontir.uity). The cracks progressed through the wall on a slightly different plane and merged about mid-wall of the valve body.
Scanning electron microscope examination of the fracture features disclosed the cracks propagated transgranularly and exhibited clearly defined grain structure striations characteristic of cyclic fatigue failure.
Cracks in the thermal sleeve and safe end sections exhibited similar fracture morphology. No evidence of corrosion interaction from chemical attack was identified.
During the design phase, Babcock and Wilcox (B&H) performed the stress analysis on the primary system up to the affected check valve which is the design code (USAS B31.7-USAS B31.1) interface boundary. Gilbert Associates, as architect-engineer, performed the balance of plant design.
The B&W design calculations for the HPI lines included a pipe section that was not installed during plant construction. The potential thermal discontinuity at this point is believed to be partly responsible for the cracking and is currently being evaluated by both organizations.
Based on the above findings, the mode of cracking was tentatively attri-buted to thermal cycle fatigue. However, the synergistic thermal-hydraulic effects contributing to the failure mechanism are yet to be detern.ined.
Contributing factors being investigated include operational design limits and setpoints with regard to makeup water temperature and flow rate, minimum bypass flow, and system thermal-hydraulic parameters around the HPI nozzle used for makeup.
B.
Oconee When the pipe u tension at Oconee 3 was removed to gain access to the thermal sleeve in order to repair it, liouid penetrant testing (LPT) disclosed cracks on the ID surfaces of the makeun/HPI pipe extension and nozzle safe end.
Crack features were similar in nature to those founo at Crystal River.
Reportedly, the cracks penetrated up to 20 ner-cent of the thickness of the pipe wall. The other makeup nozzle assembly was examined by radiography and a special ultrasonic testing (UT) tech-nioue developed by B&W for this purpose.
No indication of cracking or degraded thermal sleeve conditions was observed.
Further UT and radio-graphic testing (RT) of the two remaining HPI nozzle assemblies indicated a loose thermal sleeve in one of the nozzles (Nozzle 3B1).
At Oconee 2, results of the UT and RT indicate the thermal sleeve in one of the makeup nozzles may be loose and the retaining button welds on the safe end side are nissing. Cracking was also found in the safe end and l
Dipe extension.
The other makeuo nozzle showed no indications of a l
degraded thermal sleeve or cracking.
Examination of the two remaining HPI nozzle assemblies indicated a loose thermal sleeve (i.e., retaining weld buttons missing) in one and a crack in the rolled area of the other nozzle thermal sleeve.
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4 IN 82-09 March 31, 1982 Page 3 of 3 At Oconee 1, examination of the four HPI nozzle penetrations to the RC loop inlet line showed no evidence of degradation.
Discussion:
In B&W design plants the line(s) for nonnal makeup of reactor coolant are also part of the redundant high pressure injection system. These plants do not have a regenerative heat exchanger in the makeup coolant circuit. Therefore, during operations, the potential exists for the makeup coolant temperature to be much lower than the reactor coolant temperature in the loop.
Fluid temperature fluctuations resulting from mixing in the HPI nozzle coupled with hydraulic effects are thought to be primary contributors to the cracking problem at Crystal River and at the Oconee plants. Although the cracking location is within the scope of the LOCA (loss-of-coolant accident) safety analysis, the existence of cracking in an area not routinely included in the program of ISI represents an unacceptable challenge to system integrity.
An evaluation of the cracking problem and its resolution has been requested of the B&W Regulatory Response Group.
Pressurized-water reactor systems of the Combustion Engineering and Westinghouse designs do have a regenerative heat exchanger in the makeup coolant line which is a separate, dedicated systen.
During normal power operatfon tge makeup coolant enters the nozzle at temperatures on the order of 50 -150 F below the temperature of the reactor coolant loep respectively.
However, transients may scur in which the makeup flow rate is greater than the letdown flow rate.
Depending on the frequency and duration of these transients, the makeup coolant might not be heated to the expected temperature. Therefore, the potential may exist for large temperature fluctuations in the makeup nozzle to cause problems similar to those discussed above.
Past experience has shown similar thermal fatigue problems with nozzle-thermal sleeve assemblies in other systems of both BWR(NED0-21821,1978) and PWR (WCAP-7477 and NED0-9693-1980) designs.
This IE information notice is provided as an early notification of a potentially significant matter that is still under review by the NRC staff.
If NRC evalua-tion so indicates, further licensee action may be requested.
In the interim, we expect that licensees will review this information for applicability to their facilities.
No written response to this information notice is requested.
If you need-additional information, please contact the Regional Administrator of the appropriate NRC Regional Office.
Attachment:
Recently issued IE Information Notices
Attachment IN 82-09 March 31, 1982 RECENTLY ISSUED IE INF0PNATION NOTICES Information Date of Notice No.
Subject Issue Issued to 82-08 Check Valve Failures on 03/26/82 All power reactor Diesel Generator Engine facilities holding Cooling System an OL or CP 82-07 Inadeauate Security Screening 03/16/82 All oower reactor Programs facilities holdina an OL or CP 82-06 Failure of Steam Generator 03/12/82 All power reactor Primary Side Manway Closure facilities holding Studs an OL or CP 82-05 Increasing Frequency of 03/10/82 All power reactor Drug-Related Incidents facilities holding an OL or CP 82-04 Potential Deficiency of 03/10/82 All power reactor Certain AGASTAT E-7000 facilities holding Series Time-Delay Relays an OL or CP 82-03 Environmental Tests of 03/04/82 All power reactor Electrical Terminal Blocks facilities holding an OL or CP 82-01 Auxiliary Feedwater Pump 02/26/82 All power reactor Rev. 1 Lockout Resulting from facilities holding Westinghouse W-2 Switch an OL or CP Circuitry Modification l
80-32 Clarification of Certain 02/26/82 All facility, Rev. 1 Requirements for Exlusive-materials and l
Use Shipments of Radio-Part 50 licensees active Materials 1
82-02 Westinghouse NBFD Relay 01/27/82 All power reactor Failures in Reactor facilities holding Protection Systems at an OL or CP Certain Nuclear Power Plants 1
OL = Operating License l
CP = Construction Permit l