Information Notice No. 82-09: Cracking in Piping of Makeup Coolant Lines at B&W Plants
SSINS No.: 6835
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
March 31, 1982
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
A visual inspection inside the reactor building at this time revealed the
leak was associated with a 2 1/2-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 injection system. After the insulation was removed from the
affected valve a 140o 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 valves was removed and liquid penetrant testing (LPT) was
performed on the accessible inside diameter (ID) surfaces including 5 inches
into the 2 1/2-inch line on the inlet side of the affect valve. This
inspection disclosed an extensive network of heat-check type cracks around
the safe end ID surface. A similar conditions 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
March 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
(OD) 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
characteristics 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.
When the pipe 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) technique developed by B&W for this purpose. No indication of
cracking or degraded thermal sleeve conditions was observed. Further UT
and radiographic 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 missing. Cracking was also found in the safe end and
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.
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
(NEDO-21821, 1978) and PWR (WCAP-7477 and NEDO-9693-1980) designs.
This Information Notice No. is provided as an early notification of a
potentially significant matter that is still under review by the NRC staff.
If NRC evaluation 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.
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