ML20209J207
| ML20209J207 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/22/1986 |
| From: | Haynes J ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| ANPP-37839-JGH, DER-86-23, NUDOCS 8609160201 | |
| Download: ML20209J207 (5) | |
Text
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'M l/; y /7 Arizona Nuclear Power Project 4
P.O. BOX 52034 e PHOENIX, ARIZONA 85072-2034
, /p.
/?EC/g,,7 Y/4 August 22, 1986 ANPP-37839-JGH/DRL-92.11 4
U. S. Nuclear Regulatory Commission Region V 1450 Maria Lanc - Suite 210 Walnut Creek, CA 94596-5368 Attention:
Mr. D. F. Kirsch, Acting Director Division of Reac'.or Safety and Project Palo Verde Nuclear Generating Station (PVNGS)
Unit 3 Docket No. 50/530
Subject:
Interim Report - DER 86-23 A 50.55(c) Potentislly Reportable Deficiency Relating To A Broken Nut On The Steam Generator Sliding Base Support File: 86-019-026; D.4.33.2
Reference:
(A) Telephone Conversation between A. Hon and D. R. Larkin on July 25,1986 (Initial Notification - DER 86-23)
Dear Sir:
The NRC was notified of a potentially reportable deficiency in the ref-erenced telephone conversation.
At t hat time, it was estimated that a determination of reportability would be made within thirty (30) days.
(August 25, 1986)
Due to the extensive investigation and evaluation required, an Interim Report is attached.
It is now expected that this information will be finalized by September 30, 1986, at which time a complete report will be submitted.
Very trul yours,
/tL)$
J. G. Ilay Vice President Nuclear Production JGH/DRL:kp Attachment l
cc: See Page Two i
8609160201 860822 PDR ADOCK 05000530 S
\\ \\
-r DER 86 Interim Report Mr. D. F. Kirsch Acting Director Page Two August 22, 1986 ANPP-37839-JGH/DRL-92.11 cc:
J. M. Taylor Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C.
20555 A. C. Gehr (4141)
R. P. Zimmerman (6295)
Records Center Institute of Nuclear Power Operations 1100 Circle 75 Parkway - Suite 1500 Atlanta, Georgia 30339
INTERIM REPORT - DER 86-23 POTENTIAL REPORTABLE DEFICIENCY ARIZONA NUCLEAR POWER PROJECT PVNGS UNIT 3 I.
Potential Problem The steam generator lower support is composed of two 12 inch thick forgings which resist laterial loads, four vertical support pads to resist downward compreesive loads and eight 6 inch diameter anchor bolts to resist uplift forces.
The larger of the two forgings of the steam generator lower support is designated as forging "A";
the smaller as forging "B."
Forging "A" serves as a key to resist the translational forces transmitted by the steam generator sliding base when the steam generator is subjected to pipe break and seismic loads. This forging is anchored by twelve 5-1/2 inch diameter ASTM A540 Grade B23 Class 1 steel studs and nuts to the top of a concrete pedestal.
On the afternoon of July 11, 1986, it was discovered that an approximately one inch wide, wedge shaped piece had fallen out of a nut anchoring the Unit 3 Number 1 (Tag. No. 3 MRCE E01A) steam generator's forging "A".
The Palo Verde Project purchased the studs and nuts from Marathon Steel Company, Phoenix, Arizona.
Marathor Steel Company in turn used several subtier suppliers as sources for the studs and nuts.
A review of the Certified Material Test Report (CMTR) revealed that Jos. Dyson & Sons, Inc. of Painesville, Ohio, supplied the failed Unit 3 nut.
Bechtel Drawing 13-C-ZCS-605, Revision 11, " Containment Internals, Steam Generator Lower Supports, Sections and Details," specified high strength anchor bolts, heavy hexagonal nuts and washers meeting ASTM Specification A540 Grade B23 Class 1 (E-4340-H). The documentction search conducted at the jobsite showed the failed nut to be ASTM A540 Grade B24 Class 1.
A subsequent evaluation determined that Grade B24 is an acceptable substitute for Grade B23 material.
II.
Approach To and Status of Proposed Resolution As an initial step of the problem resolution, all (24) accessible 5-1/2 inch diameter nuts on forging "A" of both the Number 1 (Tag No. 3 MRCE E01A) and Number 2 (Tag No. 3 MECE E01B) steam generators of Unit 3 were visually examined.
The fractured nut ca the Number 1 steam generator lower support was the only nut shown to be cracked.
In addition to the visual examination, excluding the fractured nut, eighteen out of a total of twenty-three remaining 5-1/2 inch diameter nuts were accesible for hardness testing and ultrasonic examination.
Ikrdness tests using an Equotip hardness tester, det ermined the average hardness of the fractured nut to be 42.5 HRc.
The hardness of the fractured nut was verified by Rockwell C hardness measurements in the laboratory.
The average hardnesses for the eighteen accessible nuts ranged from 37.3 to 41.6 HRc.
The ultrasonic examination of eighteen accessible 5-1/2 inch diameter nuts did not detect any cracking.
In addition, ultrasonic examination of the stud associated with the f ractured nut did not detect any cracking.
L
In order to determine the nature and cause of cracking, the fractured nut was subjected to metallographic examination, chemical analysis, mechanical tests and various other tests. The metallurgical evaluation concluded:
1.
The nut failed due to stress corrosion cracking (SCC).
2.
The high hardness (43.51Rc) of the nut, water intrusion and the combined (residual and applied) hoop stresses near the nut base were the primary causes of SCC.
3.
The steel met the material specification requirements (ASTM AS40, Grade B24) except for Charpy impact values (i.e.,
16.5 ft-lbs vs.
25 ft-lbs minimum required).
IIowever, there is little correlation between Charpy impact values and susceptability to SCC.
Of major concern is the hardness of the material.
In the case of the failed
- nut, its hardness was at the high end of the acceptance range (43.5 vs. a maximum of 44.51Rc).
4.
No material flaws such as forging defects or quench cracks were found.
When considering the rule making which resulted in a recent revision to General Design Criteriou 4 for high energy line breaks in primary loop
- piping, the original design loads are reduced by approximately 80 percent.
The reduction in original design loads results in far fewer studs being required. Therefore, it can be concluded that even with the loss of the one broken nut, the support has the added capacity to perform its intended function.
A documentation review for Unit 3 was conducted to identify the location of other ASTM A540 material procured by Bechtel.
Of specific interest were certified material test reports for holts, studs end nuts that indicated material hardnesaea aqual to or reater than 41 FRc.
The review identified thc following installationst 1.
2-1/2 irt diameter heavy hex nuts for the steam generator upper support (Custom Bolt, heat trace No. SC).
l 2.
6 inch diameter steam generator lower support anchor boltr (Daido Steel. heat trace Nos. 08 and 11).
3.
3-3/4 inch diameter heavy hex nuts for the steam generator lower support, forging B (Jos. Dyson & Sons, heat trace No. C74D).
4.
2 inch diameter pressurizer anchor bolts (Daido Steel, heat trace No.
26).
I Visual examination of all nuts and anchor bolts identified with the above l
listed heat trace codes did not reveal any additionally fractured nuts or anchor bolts.
liardness testing using the Equotip hardness tester was performed on eleven
[
1 2-1/2 inch diameter nuts for the steam generator upper support and sixteen 2 inch diameter pressurizer anchor bolts.
In addition, hardness tests were performed on the sixteen 3-3/4 inch diameter nuts for the steam
(
genertor lower support.
The six inch diameter steam generator lower support tie down anchor bolts were not accessible for hardness testing.
The results of the hardness testing concluded that the f ractured nut had the highest hardness of all nuts or studs tested.
None of the additional nuts or stude tested exceeded an average of 41.51Rc.
a r,
In addition to the details given above, ultrasonic examination of all (16 total) Unit 3 six inch diameter steam generator lower support tie down anchor bolts, all (16 total) Unit 3 two inch diameter pressurizer anchor bolts and eleven Unit 3 2-1/2 inch diameter heavy hex nuts for the steam generator upper support was performed.
This ultrasonic examination did not detect any cracking.
Hardness testing did reveal another problem with material supplied by Custom Bolt Manufacturing Company.
Hardness testing resulted in identifying six 2-1/2 inch diameter heavy hex nuts for the steam generator upper support having hardness below the values specified in ASTM A540 Grade B23 Class 1.
The acceptable minimum hardness value for this material is 321 Brinnell. The hardness for the six nuts ranged from 200 to 220 Brinnell.
Subsequent chemical analysis using the Texas Nuclear Alloy Analyzer identified the six nuts as being made of carbon steel.
Two other heats of 2-1/2 inch diameter nuts supplied by Custom Bolt were identified and a random sample of each was selected.
Ten of thirty-six nuts from heat No. SB and ten of forty-three nuts from heat No. 5A were randomly selected from hardness testing The hardness testing resulted in identifying three " soft" nuts in heat No. 5B with hardness values of 195 Brinnel, 200 Brinnel and 214 Brinnel and none in heat No.
SA.
Engineering evaluation of the " soft" nuts determined that carbon steel nuts having a minimum hardness of 156 Brinnel would be acceptable.
DER 86-26 has been initiated to address the investigation and evaluation of the " soft" nut problem.
i III. Projected Completion of Corrective Action and Submittal of the Final Report It is concluded that the 5-1/2 inch diameter nut failed due to stress corrosion cracking.
Conclusions regarding Units 1 and 2 transportability are still under consideration and will depend upon the outcome of the Unit 3 investigation. Site investigation, engineering evaluation and the Final Report are forecast to be completed by September 30, 1986.
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