ML20054A443
| ML20054A443 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Cooper, Arkansas Nuclear, River Bend, Waterford, South Texas, Comanche Peak, Fort Calhoun, 05000000, Fort Saint Vrain |
| Issue date: | 03/31/1982 |
| From: | Jay Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | ARKANSAS POWER & LIGHT CO., GULF STATES UTILITIES CO., HOUSTON LIGHTING & POWER CO., KANSAS GAS & ELECTRIC CO., LOUISIANA POWER & LIGHT CO., NEBRASKA PUBLIC POWER DISTRICT, OMAHA PUBLIC POWER DISTRICT, PUBLIC SERVICE CO. OF COLORADO, TEXAS UTILITIES SERVICES, INC. |
| References | |
| NUDOCS 8204150458 | |
| Download: ML20054A443 (2) | |
Text
l
- ""*4, UNITED STATES s'
i'
)'
(
0 NUCLEAR REGULATORY COMMISSION
%/
-l REGION IV
[
l 611 RYAN PLAZA DRIVE. SUITE 1000 ARLINGTON. TEXAS 76011 March 31, 1982 Gentlemen:
The enclosed IE Information Notice provides early notification of events that may have safety significance.
It is expected that recipients will review the information notice for possible applicability to their facilities.
Sincerely, John T. Collins Regional Administrator
Enclosures:
1.
IE Infonnation Notice No. 82-09 2.
List of Recently Issued IE Information Notices N hfrS f 8 g)--
Mrm f ins s4 Akee M & z+ ce, 82041504*6 820331 PDR ADOCK 05000267 d/
[G
/
- ///
F IE INFORMATION NOTICE 8?-09:
CRACKING IN PIPING OF MAKEUP COOLANT LINES AT B&W PLANTS Licensee Facility / Docket Number Arkansas Po.<er and Light Conpany Arkansas Nuclear One, Unit 1 & 2 Little Rock, Arkansas 50-313; 50-368 Nebraska Public Power District Cooper Nuclear Station Columbus, Nebraska 50-298 Omaha Public Power District Fort Calhoun Station Omaha, Nebraska 50-285 Public Service Company of Colorado Fort St. Vrain Generating Station Denver, Colorado 50-267 Gulf States Utilities River Bend Beaumont, Texas 50-458; 50-459 Houston Lighting & Power Company South Texas Project Houston, Texas 50-498; 50-499 Kansas Gas & Electric Company Wolf Creek Wichita, Kansas STN 50-482 iouisiana Power & Light Company Waterford-3 New Orleans, Louisiana 50-382 Texas Utilities Generating Company Comanche Peak Steam Electric Station Dallas, Texas 50-445; 50-446
~
e S
RECElVED
[
~
t gpg 8 1982* 3 0.wume g -
t
.rs /%+<c 4/G
~
io :
t go,,,.
[e im JLF S4 MWF
~
Aid {b
- rn $Zo1
/E3/
M
SSINS No.:
6835 Accession No.:
8202040131 IN 82-09 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 March 31, 1982 IE INFORMATION NOTICE NO. 82-09:
CRACKING IN PIPING OF 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 2 -inch check valve (M0V-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 injec-tion system.
After the insulation was removed from the affected valve, a 140 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 on the accessible inside diameter (ID) surfaces including 5 inches into the 2 -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 side up to the disc seat area.
The valve inlet side and connecting piping were not affected.
The most severe cracking in the safe end appeared to have penetrated up to 25 percent of the wall thickness.
A visual inspection also revealed the thermal sleeve inside the high pressure injection (HPI) nozzle was loose and showed evidence of wear in areas of contact.
Some cracking of the thermal sleeve was also observed.
As a result of the Crystal River 3 findings, Duke Power Company initiated a radiographic examination of the RC inlet nozzle connections on the two HPI 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 thermal 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 extension to replace the affected thermal sleeve.
Further findings and expanded inspection as a result of this action are summarized below.
IN 82-09 March 31, 1982 Page 2 of 3 Investigation and Findings:
A.
Crystal River A m'etallurgical 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 discontinuity).
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&W) 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 determined.
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 pine extension at Oconee 3 was removed to gain access to the thermal sleeve in order to repair it, liquid penetrant testing (LPT) disclosed cracks on the ID surfaces of the makeup /HPI pipe extension and nozzle safe end.
Crack features were similar in nature to those found at Crystal River.
Reportedly, the cracks penetrated up to 20 per-cent of the thickness of the pipe wall.
The other makeup nozzle assembly was examined by radiography and a special ultrasonic testing (UT) tech-nique 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 381).
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 missing.
Cracking was also found in the safe end and
- .~
IN 82-09 March 31, 1982 Page 3 of 3 pipe extension.
The other makeup nozzle showed no indications of a 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.
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 normal 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 system.
During normal power operation, the makeup coolant enters the nozzle at temperatures on the order of 50 -150 F below the temperature of the reactor coolant loop, respectively.
However, transients may occur 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.
Enclosure IE Infor:::icn Nctice:
22-09 March 31, 1982 LISTING OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date Notice No.
Subject Issued Issued To 82-02 Westinghouse NBFD Relay 1/27/82 All power reactor facilities Failures in Reactor Protection holding an OL or CP Systems at Certain Nuclear Power Plants 80-32 Clarification of Certain 2/12/82 All fuel facility, materials, Rev. 1 Requirements for Exclusive-and Part 50 licensees Use Shipments of Radioactive Materials 82-01 Auxiliary Feedwater Pump 2/26/82 All power reactor facilities Rev. 1 Lockout Resulting from holding an OL or CP Westinghouse W-2 Switch Circuit Modification 82-03 Environmental Tests of 3/04/82 All power reactor facilities Electrical Terminal Blocks holding an OL or CP 82-04 Potential Deficiency of 3/10/82 All power reactor facilities Certain AGASTAT E-7000 holding an OL or CP 82-05 Increasing Frequency of 3/10/82 All power reactor facilities Drug-Related Incidents holding an OL or CP 82-06 Failure of Steam Generator 3/12/82 All power reactor facilities Primary Side Manway Closure holding an OL or CP Studs 82-07 Inadequate Security Screening 3/16/82 All power reactor facilities Programs holding an OL or CP l
82-08 Check Valve Failures on 3/26/82 All power reactor facilities Diesel Generator Engine holding an OL or CP Cool.ing System OL = Operating License CP = Construction Permit l
l L