ML20003B692
| ML20003B692 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/18/1981 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 11334, NUDOCS 8102250209 | |
| Download: ML20003B692 (4) | |
Text
.
O
$03CUiU3I3 M",/3r James w cook l%aQdif0,f nce Pressdent - Projects, Engsneemg and Constructson z
General offices: 1945 West Parnell Road, Jackson, MI 49201 * (517) 788 0453 February 18, 1981 Harold R Denton Office of Nuclear Reactor Regulation Nuclear Regulatory Commission Washington, DC 20555 MIDLAND PROJECT MIDLAND DOCKET NOS 50-329, 50-330 REQUEST FOR ADDITIONAL INFCRMATION ON REACTOR VESSEL SUPPORT DESIGN RESPONSE TO QUESTION 121.33 FILE:
B5.4.13, 0505.803 UFI: 02300(S), 70*01*10*03 SERIAL:
11334 Mr R L Tedesco's correspondence of September 19, 1980 forwarded the NRC Materials Engineering Branch Question 121.33 relating to our Preliminary Report No 1, July 1980, on the reactor vessel support modification.
The attached Enclosure is a complete response to this three-part question.
This written reply will document our verbal responses to parts 1 and 2 which were provided to the NRC Region III Office on August 20, 1980.
We are providing this separate response to allow the Office of Nuclear Reactor Regulati,on (NRR) to complete its review and provide the NRC Region III Office y,ith its. approval of the design concept. The NRR's question and our response
[will be,facorporated into the Midland FSAR by amendment.
S b,
JWC/RLT/qr
/
n ENCLOSURE: MATERIALS ENGINEERING BRANCH QUESTION 121.33 CC Director of Office of Inspection & Enforcement At Mr Victor Stello, USNRC w/a f
Director, Office of Management J
Information and Program Control, USNRC w/a f
Commander, Naval Surface Weapons Center At Mr P C Huang w/a oc0281-0238a100 8102 250 Roq i
o 2
Energy Technology Engineering Center Facility Design Engineering At Mr L J Auge, Manager w/a RLBaker, (B&W/AA) w/a JWCook, P-26-336B w/a RJCook, Midland Resident Inspector w/a LHCurtis, Bechtel w/a DFJudd, B&W w/a GSKeeley, P-14-113B w/a CBechhoefer, ASLB w/o GALinenberger, ASLB w/o FPCowan, ASLB w/o AS&L Appeal Panel w/o MMCherry, Esq w/o MSinclair w/o CRStephens, USNRC w/o WDPaton, Esq, USNRC w/o FJKelly, Esq, Attorney General w/o GTTaylor, Esq, Asst Attorney General w/o WIDiarshall w/o i
oc0281-0238a100 9-y.f-g 9
r*
es
'T
ENCLOSURE NRC QUESTION 121.0 MATERIALS ENGINEERING BRANCH Question 121.33: Provide the following information regarding your report (5.3, 3.9.3)
" Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Midland, Michigan, Preliminary Report No 1",
July 1980, 1.
Describe the steel-making practice used for the material of the three failed reactor vessel anchor bolts. State whether the material was aluminum or silicon killed and vanadium grain refined.
2.
Identify the material of construction, including fabrication details, to be used for the modifications to the reactor vessel support design.
3.
Describe the preservice and inservice inspection program to be conducted for the reactor vessel support design, including anchor bolts, shear pins and proposed lateral support design.
Identify all exemptions, if any, which will be needed for the support design. Discuss any provisions included to maintain radiological exposures during these inspections to as low as is reasonably achievable, and provide the predicted radiological exposure results for these inspections.
Response
1.
For the three failed reactor vessel anchor studs, the material was silicon killed and aluminum grain refined.
Vanadium was not used for grain refinement.
2.
The modifications to the reactor vessel support design will consist of strengthening the existing primary shield plug support brackets by welding stiffener plates to the existing supports and by adding a stainless steel shim block between the strengthened primary shield plug support bracket (ie, the upper lateral support) and the reactor vessel. The existing primary shield plug support bracket is a welded fabrication made from ASTM-A-516 Grade 70 plate material. These supports will be strengthened by shielded metal-arc welding ASTM-A-516 Grade 70 plates to the existing supports using E-7018 filler metal. The stainless steel shim block will be ASTM-A-240 Type XM-19 material. This shim block will be bolted to the strengthened primary shield plug support bracket.with ASTM-A-325 bolts. Please refer to Figures 4, 5 and 6 from our Report No 2 dated December 1980.
3.
The preservice and inservice inspection program for the reactor vessel supports will be in accordance with Section XI of the ASME Boiler.and Pressure Vessel Code as described in the "Preservice and First 10-Year Interval Inspection Plan for Nondestructive Examination miO281-0239a100 e
+e E8
.Q&A 121.33 2
and System Pressure Testing of Consumers Power Company Midland Plant Units 1 and 2".
This inspection plan is referenced in Section 5.2.4 of the Midland FSAR and is currently undergoing revision, but it will describe the preservice and inservice inspection programs relative to these components.
It will define the version (edition and addenda) of the ASME Code which is applicable to these components.
Following are the features of the referenced document which will relate to the inspection of the cooponents in question.
a.
Upper Lateral Supports -
Other than consturction inspections, no preservice or inservice inspections will be performed on the upper lateral supports.
This position is consistent with the inspection of other non-integral supports of the Midland Plant. The Section XI Code rules applicable to the inspection are contained in Subsection IWF of the Code and IWF-1100 (a) exempts cc=ponents such as the upper lateral supports from preservice and inservice inspections.
b.
Anchor Bolts and Shear Pins -
Here again, the applicable preservice and inservice inspection requirements are contained in Subsection IWF of the ASME Section XI Code. The anchor bolts and shear pins will be visually (method VT-3) examined in accordance with the requirements for Item F-1 in Table IVF-2500-2 of the ASME Section XI Code.
As of this time, the predicted level of radiological exposures during inservice inspections has not been fully defined, however, it has been decided to examine the anchor bolts and shear pins with a remote visual examination systen. Design of the examination system is not complete, but it will be designed to maintain radiological sxposures to as low as reasonable achievable. Si tee the inservice inspection of the bolts and shear pins would have been required by the ASME 3ection XI Code for the original reactor pressure vessal support system, there will be no additional radiological exposure resulting from the modification of the support system.
miO281-0239a100 m
_