IR 05000329/1980008
| ML19309H364 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/17/1980 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19309H356 | List: |
| References | |
| 50-329-80-08-01, 50-329-80-8-1, 50-330-80-08, 50-330-80-8, NUDOCS 8005130101 | |
| Download: ML19309H364 (6) | |
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8005130 ton, U.S. NUCLEAR REGULATORY COMMISSION
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OFFICE OF INSPECTION AND ENFORCEMENT
REGION III
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Report No. 50-329/80-08; 50-330/80-08 Docket No. 50-329; 50-330 License No. CPPR-81; CPPR-82 Licensee: Consumers Power Company 1945 West Parnall Road Jackson, MI 49201 Facility Name: Hidland Nuclear Power Plant, Units 1 and 2 Inspection At: Midland Site, Midland, Michigan Inspection Conducted: January 1-31, 1980
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Inspector:
R. J. Cook l
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Approved By:
R. C. Knop, Chief, 3-/7-[d Projects Section 1 Inspection Summary Inspection on January 1-31, 1980 Report No. 50-329/80-08; 50-330/80-08)
Areas Inspected: Pull testing of grouted anchor bolts, failure of Unit I reactor vessel anchor bolts and associated hardness testing, examination of site conditions, assembly of Unit I reactor vessel internals, presence of fluid found in 350 MCM-3/C-B-11 power cables, and corporate management meeting.
This inspection effort involved a total of 50 inspector-hours by one NRC inspector.
Results: No items of noncompliance or deviations were identified.
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DETAILS Persons Contacted Consumers Power Company Personnel D. Miller, Site Manager
- J. Corley, QA Section Head, IE & TV B. Peck, Construction Supervisor M. Shaeffer, Group Supervisor D. Keating, QA Group Supervisor P. Kyner, QA Electrical Supervisor B. Horn, Group Supervisor QAE
- E. Evans, Field Electrical Engineer
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Bechtel Power Corporation Personnel
- W. Willman, QC Engineer Numerous other principal staff and other personnel were contacted during the recording period.
- Denotes those present during the exit interview conducted during the reporting period.
Licensee Action on Previous Inspection Findings (0 pen) Unresolved Items No. (329/78-19-05; 330/78-19-05):
During the reporting period the Resident Inspector witnessed pull tests on four grouted anchor bolts which had exhibited a voided area around the ex-posed bolt ends. The pull tests were conducted as per a procedure documented on Bechtel Corporation Field Engineer's Report Form dated December 20, 1979 signed by Mr. J. Betts. The bolts tested were nomi -
nally 5/8" diameter. Specification No. 7220-C-306 Revision 3 states that, "5/8" diameter anchor bolts in 4,000 psi concrete should with-stand a 2.2 kips longitudinal load and withstand a similar 2.7 kips load when used in 5,000 psi concrete. Bolts for hangers designated 3-1HPC-145-H3 (2 bolts) and 2 1/2 - ICCB-2-H4, sketch nos. 616-4-11 and 603-7-14 respectively were pulled at 2.2 kips using hydraulic piston no. JSH-123-0A6 and held for one minute. One bolt for hanger designated 12-2GCB-25-H8 (one bolt), sketch no. 611-3-29 was pulled at 2.7 kips and held for one minute using the same hydraulic cylinder.
The area at the hydraulic piston used was certified to be 2.761
.003 square inches. The pressures maintained for load tests of anchors in 4,000 psi concrete were 800 psi and 1,000 psi for the anchors used on 5,000 psi concrete. Calibrated gauge no. N-4 was used for determining the hydraulic pressure. The gauge pressure calibrated on September 21,
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1979 by Indiand Ryerson Construction Products Company. After the pull test, each of the bolts was examined for any bolt pull-out and/or con-crete spaulding. None was detected.
Reportable Deficiencies - 50.55(e) Item Failure of Reactor Vessel Anchor Bolts As a result of spontaneous failure of reactor vessel anchor studs no.
3 (inside) and no. 36 (outside) for Unit I which were discovered on September 14, 1979 and December 19, 1979, Teledyne Engineering Services was contracted to evaluate the cause of these failures. As a part of this evaluation, Teledyne Engineering Services, performed hardness tests of all the installed intact anchor studs for both Unit 1 and Unit
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2 reactor vessels. There were originally 96 studs for each reactor vessel. Teledyne Engineering Services used an Equo-Tip Portable Hard-ness Tester for these measurements.
A cursory review of the data revealed that six bolts of Unit I were within the specification tolerance of 32-38 R Dc Q and 18 bolts in Unit 2 show hardness values greater than 38 R Dc Q.
In addition to the reactor vessels anchor studs, Teledyne Engineering Services performed hardness tests on five steam generator anchor bolts of each steam generator for both units. Each unit has two steam genera-tors. The steam generator studs tested showed localized hardness values between nominally 28.5 and 42.5 R Dc Q.
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Stress corrosion cracking is attributed as being the most probably cause for spontaneous failure with a small critical flaw size.
Regional based inspectors are following this matter.
Functional or Program Areas Inspected 1.
Site Tours At periodic intervals during the report period, tours of randomly selected areas of the site were performed. These tours were in '
tended to assess the cleanliness of the site; storage conditions of equipment and piping being used in site construction; the potential for fire or other hazards which might have a deleterious effect on personnel and equipment and to witness construction activities in progress.
It was noted that the licensee was improving control of extraneous welding materials such as welding stubs and placement of five blankets.
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Unit 1 Reactor Vessel Internal Assembly During the reporting period the Resident Inspector examined and witnessed selected portions of fitting the core support assembly guide blocks. B&W was having some difficulty in maintaining the 0.03 inch levelness requirements specified in B&W Procedure No.
FCP-132, Initial Fit-Up of Internals Core Support Assembly, when the core support assembly was engaged with the reactor vessel
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alignment keys. This is partially attributed to the reactor vessel
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being set within drawing levelness requirements but not within the 0.003 inch levelness requirement between parallel mating surfaces of the core support assembly and the reactor vessel. Consumers Power Company generated Nonconformance Report (NCR) No. M-0}-4-0-004 addressing the out of alignment condition. B&W generated a Site Problem Report (SPR) No. 13-12-144-00 which also addressed the out of alignment condition. The response to the B&W SPR indicated that 0.005 to 0.006 inch measured misalignment with the core support assembly engaged with the reactor vessel alignment keys was accept-able and were continued on further fit-up of the guide blocks.
During the guide block fitting evolution, communications were established between Consumers Power Company and the B&W factions.
However, it could not be determined whether actual or implied pro-cedural requirements (as written and used) were actually violated.
Therefore, this is considered an unresolved item pending further review. Unresolved Item (329/80-08-01).
The assembly of the internals lifting device was noted during these evolutions and compared to B&W Drawings 203275E and 203254E with particular emphasis on placement of safety wires. The pendants used in Unit I were those used for Unit 2 with the settings adjusted for Unit 1.
The pendants for Unit I were not available at this time.
Final pendant adjustment will be required for both Units 1 and 2 at a later time. After use, the internals lifting device was wrapped with polyethylene and stored in a protected area of the containment building. No discrepancies were noted.
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Presence of Fluid in 350 MCM - 3/C B-11 Power Cable As discussed in NRC Inspection Report No. 50-329/79-13, 79-15, 79-27; j
50-330/79-13, 79-15, 79-27, the licensee has received 350 MCM Power Cable from Essex which has a fluid substance in one phase. Discus-
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sions with the licensee during the report period revealed that addi-
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tional testing and evaluation of the suspect cable is still being perfo rmed. No conclusions as to the usefullness of the cable have been reached at this time.
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Management Meeting On January 11, 1980, the Resident Inspector attended a meeting along with other regional personnel, with Consumers Power Company management at their corporate offices located in Jackson, Michigan. Construction matters requiring resolution before issuance of an operating license; NRC inspection program as it relates to the Pre-operational and Power Assention Programs and understanding of the licensee's program for staffing, start-up, and operations of the Midland facility were discussed during the meeting. NRC Inspection Report No. 50-329/80-02; 50-330/80-02 references this meeting in more explicit detail.
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Unresolved Matters
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Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of.non-compliance or deviations. One unresolved item, disclosed during this inspection, is discussed in paragraph 2.
Exit Interview The Resident Inspector met with licensee representatives (denoted under Persons Contacted) on January 17, 1980. The inspector summarized the scope and findings of the inspection effort to date. The licensee acknowledged the findings reported herein.
Attachment: Preliminary Inspection Findings During fitting of the core support assembly guide blocks for Unit 1, it could not be determined whether procedural requirements (as written and j
used) were actually violated for B&W Procedure No. FCP-132, initial fit-up of internals core support assembly, steps 260 through 360.
This item is considered unresolved pending further review.
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M C.G Su 2. REGIONAL OFFICE
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Consumers Power Company
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1945 West Parnall Road-
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Jackson, M1 49201
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Office of Inspection 4 Enforcement, RI::..'
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Midland Unit 1
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DOCXCT N"M32RS
LICENSE N'232RS 5.
DATE OF INSPECTION
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50-329 50-330..,
CPPR-81 CPPR-82 N -
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h, 7.. ~,*. The '. foll.owing" matters ' ate. pre. l'indni.ry'.*1ns. pe'e.tio'n / fin. ding's s.'
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DUAlrJG i'ITTIblG * c9 THC. Code SUffC/lT ASSEM8Lf 6 Utbc blocks
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Theseprelininaryinspectionfindingavf.11'be'reviewedby'NR5"Sdpedi5 ton /.~l,1
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Manageacn at the Region 11: Office and they Trill correspond with you
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