ML20215F464
| ML20215F464 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/25/1986 |
| From: | Haynes J ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| ANPP-38430-JGH, DER-86-26, NUDOCS 8610160216 | |
| Download: ML20215F464 (5) | |
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- EP Arizona Nuclear Power Project
/Sg P.O. BOX 52034 e PHOENIX ARIZONA 85072-2034 M
If September 25, 1986 ANPP-38430-JGH/LAS
'-92@l U. S. Nuclear Regulatory Commission Region V 1450 Maria Lane - Suite 210 Walnut Creek, California 94596-5368 Attention:
Mr. D. F. Kirsch, Acting Director Divison of Reactor Safety and Projects Palo Verde Nuclear Generating Station (PVNGS)
Units 3 Docket Nos. 50/530
Subject:
Interim Report - DER 86-26 A 50.55(e) Potentially Repertable Deficiency Relating To Sof t Nuts on Steam Generator Upper Supports File: 86-006-216; D.4.33.2
Reference:
Telephone Conversation between R. C. Sorenson and D. R. Larkin August 27,1986 (Initial Notification - DER 86-26)
Dear Sir:
The NRC was notified of a potentially reportable deficiency in the referenced telephone conversation. At that time, it was estimated that a determination of reportability would be made within thirty (30) days. (September 26, 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 October 30, 1986, at which time a complete report will be submitted.
Very trul yours, M
J. G. Hayhes Vice President Nuclear Production JGH/DRL:kp Attachments cc: See Page 2 8610160216 86092S DR ADOCK 0500 0
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DER 86 Interim Report Mr. D. F. Kirsch Acting Director Page Two September 25, 1986 ANPP-38430-JGH/LAS/DRL-92.11 cc:
J. M. Taylor Office of. Inspection and Enforcement U. S. Nucicar Regulatory Commission Washington, L. 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
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INTERIM REPORT - DER 86-26 POTENTIAL REPORTABLE DEFICIENCY ARIZONA NUCLEAR POWER PROJECT PVNGS UNIT 3 1.
Potential Problem The upper steam generator supports consist of three parts; the upper keyways, lever supports and snubber support assemblies.
The upper keyways (located north and south of each steam generator) are composed of two ASTM A336 forgings located on either cide of the steam generator key. Each forging is bolted by twelve,1-1/2 inch diameter, preloaded ASTM A540 Crade B23 Class 1 studs and nuts.
These forgings resist the thermal, seismic, primary loop and main steam line pipe break loads transmitted by the key.
l During Equotip hardness testing to resolve DER 86-23, six 2-1/2 inch diameter heavy hex nuts holding the Unit 3, Steam Generator Number 1, canal wall keyway forgings to the embed were found to have hardnesses less than the minimum specified for ASTM A540 Grade B23 Class 1 bolting material. Hardness of the six nuts ranged frcm 200 to 220 Brinnell whereas the minimum required by the material specification is 321 Brinnell.
The Palo Verde Project purchased the nuts from Marathon Steel Company, Phoenix, Arizona. Marathon Steel Company in turn used several subtier suppliers as sources for the nuts. A review of the Certified Material Test Reports (CMTR) revealed that Custom Bolt Manufacturing Company supplied the " soft" nuts.
Bechtel Drawing 13-C-ZCS-606, Revision 7, " Containment Internals, Steam Generator Upper Supports, Sections and Details, specified high strength anchor bolts, heavy hexagon nuts and washers meeting ASTM Specification A540, Grade B23, Class 1 (E-4340-H).
II.
Approach To and Status of Proposed Resolution As part of DER 86-23's resolution plan, a review of certified-material' test reports of ASTM A540 material was conducted. The purpose of this review was to identify bolts, studs and nuts in which the CMTR indicated that material hardness was 41 HRC or greater. This review identified 2-1/2 inch diameter heavy hex nuts for the steam generator upper support marked with heat trace code SC.
The resolution plan for DER 86-23 required that the Unit 3 nuts be visually examined, hardness tested and ultrasonic tested to determine if there was any evidence of stress corrosion cracking.
Visual examination of all accessible 2-1/2 inch diameter nuts for the Unit 3 steam generator upper supports was performed. Only eleven 2-1/2 inch diameter nuts with heat trace marking SC could be located in the Unit 3 upper steam generator supports. All eleven of these nuts were subjected to visual, hardness and ultrasonic testing. No evidence of stress corrosion cracking was found, however, the Equotip hardness testing revealed that six of the eleven nuts had hardness less than that required by the ASTM material specification.
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Subsequent to hardness tests a review was made to identify all ASTM A540 material supplied by Custem Bolt Manufacturing Company through Marathon Steel. The review determined that Custom Bolt supplied j'
only non-ASME material to Marathon Steel for the project. Specific applications of Custom Bolt material were as follows:
a.
2-1/2 inch diameter AS40 Grade B23 Class 1 heavy hex nuts for the steam generator upper support-(heat trace codes 5A, 5B, and SC).
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b.
1-1/4 inch diameter A540 Grade B23 Class 1 anchor bolts for the steam generator upper snubber support assembly anchorage (heat numbers 46014 and 88078).
c.
1-1/4 inch diameter A540 Grade B23 Class 1 heavy hex nuts for the steam generator upper snubber support assembly anchorage (heat number 8071070).
Visual inspection of the Unit 31-1/4 inch diameter studs and nuts failed to locate material supplied by Custom Bolt.
Inspection of the Unit 3 steam generator upper supports did locate 36 2-1/2 inch -
diameter nuts of Custom bolt heat trace code 5B and 43 2-1/2 inch diameter nuts with heat trace code SA.
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In determining if 2-1/2 inch diameter nuts with heat trace codes 5A and 5B were also involved, ten nuts of each heat in Unit 3 were randomly selected. Hardness testing of the randomly selected nuts identified three nuts from heat trace code 5B that had hardness less than required. The hardness of all nuts tested with heat trace code i
5A met specification requirements. Thus a total of nine 2-1/2 inch diameter nuts (six from No. SC and three from No. 5B) had hardness less than the specification requirement. Hardness of the nine
" soft" nuts ranged from 199 to 220 Brinnell.
All nine "sof t " nuts were then subjected to in-situ chemical i
analysis using the Texas Nuclear Analyzer. The chemical analysis revealed that the nine " soft" nuts were of carbon steel. Thus it was concluded that the "sof t" nuts resulted from a material mix-up by the nut supplier rather than deficiencies in the heat treatment.
L Engineering evaluation of the "sof t" nuts determined that carbon steel nuts with a minimum hardness of 156 Brinnell (78 Ksi tensile strength) would meet strength requirements. A calculation (Calc.
j-No.13-CC-ZC-141, Rev. 8) was performed for both the 2-1/2 inch diameter bolting for the upper steam generator keyway forgings and the 1-1/4 inch diameter bolting for the steam generator snubber support assembly anchorage. The calculations assumed that all the nuts were carbon steel and the studs were A540 material. The assumption that the studs were of A540 material is valid since all studs were successfully preloaded. Had the material been carbon steel, the studs would have yielded resulting in deformation of the threads such that the nuts could not be run down.
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As a part of the engineering evaluation, a statistical analysis of the nine hardness values for the " soft" nuts was made. The statistical analysis revealed that with a 95 percent confidence level no more than 0.1 percent of the tota 1' population (480 nuts for three units; 288 for the keyway, 192 for the lever supports) would
' have a hardness less than 167 Brinnell. Thus it is expected that
-i' none of the 2-1/2 inch diameter nuts would have a hardness less than the 156 Brinne11' required by design.
Although the sample population for the 2-1/2 inch diameter nuts is limited to Unit 3, hardness testing of nuts from Unit 1 and 2 is not required because:
a.
Engineering calculations showed carbon steel nuts with a 156 Brinnell as meeting strength requirements.
b.
The high confidence level of not having material with a Brinnell hardness less than 167.
c.
Review of CIP's determined that all studs have been successfully pre-tensioned using the direct tensioning method. The i
pre-tension. loads would be the greatest load the studs and nuts would be subjected to.
Therefore, they have all been proof tested.
When considering the 1-1/4 inch diameter nuts and studs from Custom Bolt, hardness testing is not required because of reasons (a) and (c) given above'except that_ some of the 1-1/4 inch studs were j
pretensioned using the turn-of-the-nut method._ From a dccument review, it is believed that 261-1/4 inch ' studs were installed in Unit.1.
In addition, records show that the project received 192 1-1/4 inch diameter heavy hex nuts from _Custem: Bolt. It is. believed
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that 96 of'the nuts are embedded in concrete and distributed among the three units.
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III. Projected Completion of Corrective Action and Submittal of the Final Report It is concluded that a material mix-up in the supplier's shop resulted in a number of carbon steel 2-1/2 inch diameter heavy hex l-nuts being represented as ASTM AS40 Grade B23 Class 1 material.
1 Conclusions regarding transportability in Unit 1 and 2 is still under investigation. Site investigation, engineering evaluation and the Final Report are forecast to be completed by October 30, 1986.
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