IR 05000315/1987023
| ML17325A341 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/16/1987 |
| From: | Danielson D, James Gavula, Jeffrey Jacobson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17325A340 | List: |
| References | |
| 50-315-87-23, 50-316-87-23, NUDOCS 8710260005 | |
| Download: ML17325A341 (7) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/87023(DRS);
50-316/87023(DRS)
Docket Nos.
50-315; 50-316 Licenses No.
American Electric Power Service Corporation Indiana and Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Facility Name:
D.C.
Cook Nuclear Plant, Units 1 and
Inspection At:
D. C.
Cook Site, Bridgman, Michigan Inspection Conducted:
August 20, September 10, 15, and 30, 1987 Inspectors J.
M. Jacobson Da e
J.
A. Gavula
/o 48+
Date Approved By:
D.
H. Danielson, Chief Materials and Processes Section JN/g +
Date Ins ection Summar Ins ection on Au ust 20 Se tember
15 and
1987 (Re orts No. 50-315/87023(DRS)
50-316 87023(DRS System piping failures (92705), the containment hatch cover bolting discrepancies (92700),
and participate in an NRR meeting with the licensee regarding the upcoming steam generator replacement (94702).
Results:
No violations or deviations were identified.
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Persons Contacted DETAILS American Electric Power Service Com an (AEPSC)
Indiana and Michi an Power Com an 18MPC)
"+M. P. Alexich, Vice President, Nuclear
"+C.
A. Erikson, Assistant Division Manager, NOD
- +V. A. Lepore, Manager, Design Division
- +W.
P.
Modry, Supervising Designer, Structural Design
+J.
A. Reihiger, Design Engineer
"+P.
A. Barrett, Senior Engineering, NS8L
+T.
R.
Satyan-Sharma, Senior Engineer, NSKL
+D.
A. Patience, Staff Metallurgist
'R.
F. Hanlon, Design Engineer
~R.
A. Snyder, Civil Engineer
'C.
S.
Togsa, Section Manager
'T.
G. Harshbarger, Licensing S. J.
Brewer, Section Manager
"N. Ruccia, Structural Design
"B.
G.
Sheares, Civil Engineer
- F.
L. Lewis, Civil Engineer
"J.
A. McElligoit, Supervisory Auditor
"E. Morse, NOE Supervisor
"J.
Sampson, Supervisor of Safety and Assessment USNRC QOyJ
+OyJ oD.
oF oT oD M. Jacobson, Region III, Reactor Inspector A. Gavula, Region III Reactor Inspector C. Jeng, NRR, Section Chief Rinaldi, NRR/ESGB, Engineer Alexon, NRR/PD III-3, Project Manager Wigginton, NRR/PD III-3, Project Manager
+Attended September 10, 1987 meeting at D.C.
Cook Site.
'Attended September 15, 1987 meeting at NRR Headquarters.
"Attended September 30, 1987 meeting at D.C.
Cook Site.
Containment Hatch Cover Boltin Deficiencies A meeting was held at the site on September 10, 1987, to discuss the recently identified equipment hatch cover bolting deficiencies.
According to the licensee, these deficiencies occurred during original construction, although no definite record had been found.
During the meeting, the licensee presented the results of the Unit 1 RCP and CROM hatches.
These calculations assumed that any deficient bolt did not provide support to the hatch covers during design basis 1'oading conditions.
Under these conservative assumptions, the stresses and displacements of the as-found configuration met the necessary design criteria.
On this basis, the Unit 1 hatch covers were considered to have been operabl ~
At the time of the meeting, the analyses for the Unit 2 hatch bolt deficiencies were not completed.
Due to the extensive nature of the discrepancies, additional analyses were necessary.
As a followup to the September 10, 1987, meeting, members of the NRC Regional Staff met with licensee representatives at the site on September 30, 1987, te discuss the Unit 2 equipment hatch cover bolting deficiencies.
During the meeting, the licensee presented the results of the analyses for the as-found condition of the Unit 2 RCP hatch covers.
The analyses utilized the ultimate bolt capacities that were determined by in situ testing.
Using these actual values in elastic and non-linear analyses, it was determined that the stresses and displacements of the various as-found configurations met all the necessary design criteria.
On this basis, the Unit 2 hat'ch covers were considered to have been operable.
Also during the September 30, 1987, meeting, the licensee presented their proposed modification for all of the deficient hatch bolts on Unit 2.
The intent of this modification was to return the deficient bolts to their original design capacity.
The NRC inspectors had no adverse comments concerning this information.
Although it was shown that the deficient bolt configurations did not affect the operability of the RCP and CROM hatch covers, the question of how these deficiencies occurred must still be answered.
Pending the review of applicable design documentation and the determination of loss of configuration control during original construction, this is considered an Unresolved Item.
(315/87023-01; 316/87023-01)
3.
Steam Generator Re lacement Meetin The licensee met with members of the NRC staff from NRR and Region III in Bethesda, Naryland on September 15, 1987, to discuss several issues for the upcoming steam generator replacement project.
The first topic was the proposed modifications to the auxiliary building crane girders to upgrade the existing 150 ton Category I crane to a 300-ton single failure proof crane.
The other topic involved the sampling plan and visual inspections for the cadweld splices on the steam generator enclosure rebars.
Both of these topics will be dispositioned by the NRC's NRR staff.
Com onent Coolin Water Pi in Failures A through-wall crack on the 14" diameter Component Cooling Mater (CCM)
return line from the Unit 2 RHR heat exchanger was found during a routine inspection by plant personnel.
The CCM system is safety-related and as such.the licensee decided to further evaluate the system integrity.
The CCW system is closed loop and provides. cooling water to heat exchangers, pump seals, etc; The system design pressure and temperature are 150 psig and 60'.to 120'
respectively.
The piping is carbon steel seamless construction conforming to ASTM A-106, Grade B.
For pipe diameters
greater than 2" diameter, weld joints are butt weld utilizing backing rings.
For piping 2" diameter and less, socket weld fittings are used.
The water in the CCW system is demineralized water treated with 1000 ppm of sodium nitrite and adjusted to a pH of 8.5-9.0 with sodium hydroxide.
The water had been stagnant in many of the system lines much of the time.
The licensee, upon further inspection of the system, found a total of 14 through-wall'cracks and six subsurface cracks on Unit 2.
The NDE performed consisted of magnetic particle examination on the welds 2" diameter and under, and a combination of ultrasonic and radiographic examination of welds greater than 2" in diameter.
A total sample of 448 welds encompassing safety train services, important to operation, and balance of service categories were examined.
The NRC inspector reviewed the following radiographs and found the licensee's interpretation to be conservative in nature.
2CCW-131, Sh 3, W-11 2CCM-45, Sh 2, W-5 2CCW-131, Sh 2,
.W-16 2CCW-88, W-4A 2CCW-13, W-9 2CCW-20, W-6 2CCW-19, W-8 2CCW-20, M-7 2CCW-81, W-15 2CCW"81',
W"13 2CCW"41, W-39 The Unit 1 sample inspection included approximately 80 welds with a total of four through-wall cracks discovered at small bore pipe welds.
The licensee is currently developing an NDE program for future examination of CCW welds.
The review of this program is considered an Open Item (315/87023-01; 316/87023-01)
Metallurgical samples of the cracked piping were examined by both Gelles Labs, Inc.
and Westinghouse.
Examinations included spectrographic analysis, scanning electron microscopy, light microscopy, energy dispersive x-ray analysis (EDS), wavelength dispersive analysis (WDS), radiography, and magnetic particle examination.
The spectrographic analysis confirmed the pipe samples met the standard composition limits for the specified materials.
Microscopy revealed some cracking to be intergranular in nature and caused by a stress corrosion related mechanism.
Both the EDS and WDS examinations failed to identify the corrodent.
The NRC inspector reviewed the metallurgical studies and found the methodology and conclusions to be acceptable.
Westinghouse was contracted by the licensee to perform a fracture mechanics analysis to determine the maximum length of through-wall crack in each pipe diameter which would allow continued operation without jeopardizing safety.
Loadings consisting of pressure, deadweight, thermal expansion and seismic were considered in the analysis.
The approach used followed the guidelines of the flaw evaluation procedure recently developed for ASME Section XI.
Both the global instability and the locally dominated J integral instability were considered.
Leak rate predictions for given through-wall crack sizes were also considered.
The NRC inspector reviewed the Mestinghouse analysis and found the methodology and calculations to be acceptabl \\
The licensee has repaired all known through-wall cracks and has used the Westinghouse fracture mechanics analysis to evaluate existing subsurface cracks.
The two most likely scenarios for initiating the cracking are nitrite depletion in the area of a crevice or nitrite transformation into nitrate as a result of bacterial action.
The licensee has included a
biocide (Gluteraldehyde)
in the water chemistry and increased the target concentration of nitrite as per the Westinghouse recommendations.
These actions are thought to arrest the initiation of additional cracks.
The known subsurface cracks will be monitored using NDE methods.
Also, it is the NRC inspectors understanding that th'e portion of CCW piping containing the two-inch by five-inch cutout/repair will be replaced during the upcoming outage.
5.
~0en Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involves some action on the part of the NRC or licensee or both.
Open items disclosed during this inspection are discussed in Paragraph 4.
6.
Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an. open item, a
deviation, or a violation.
Unresolved items disclosed during this inspection are discussed in Paragraph 2.
7.
Exit Interview The Region III inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection.
The inspectors summarized the purpose and findings of the inspection.
The licensee representatives acknowledged this information.
The inspectors also discussed the likely informational content of the inspection report with regards to documents or processes reviewed during the inspection.
The licensee representatives did not identify any such documents/processes as proprietary.