IR 05000334/1978030
| ML19259A686 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 12/01/1978 |
| From: | Keimig R, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19259A685 | List: |
| References | |
| 50-334-78-30, NUDOCS 7901100088 | |
| Download: ML19259A686 (5) | |
Text
'
~
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFCRCEMENT Region I Report No.
7A_10 Docket No.
50-334 License No.
DPR-66 Priority Category C
-
Licensee:
Duquesne Light Comoany 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Facility Name:
Beaver Valley Power Station, Unit 1 Inspection at:
Shippingport, Pennsylvania Inspection conducted: October 31 - November 3,1978 Inspectors:
J
/-f, i
'd//7/
/2 c9 W.J.Raymond(Reacto'rInspector date signed date signed
-
date signed
,,
C
/2*/~ Approved by: s . R. Kei , Chief, actor Projects date signed Section .1, RO&B ranch Inspection Summary: Inspection on October 31 - November 3,1978 (Report No. 50-334/78-30) Areas Inspecteo: Routine, unannounced inspection by a regional based inspector of BVPS fire protection program; license event report on SI system piping supports; and review of Unit 2 containment liner weld problems for applicability to Unit 1.
The inspection involved 25 inspector hours on site by an NRC region-based inspector.
Results: No items of noncompliance were identified.
12790110o ORE Region I Form (Rev. April 77)
. s DETAILS 1.
Persons Contacted Duquesne Light Company Mr. J. Carey, Technical Assistant - Nuclear
- Mr. J. Hrivnak, Station QA Mr. A. Mosso, QA - NDE Specialist Mr. W. Sikorski, QA - NDE Supervisor Mr. R. Swiderski, Superintendent of Construction
- Mr. H. Williams, Chief Engineer Stone and Webster Corporation Mr. P. Ward, Lead Engineer
- denotes those present at exit interview 2.
Fire Protection Program Review . The inspector accompanied the fire protection program review team during inspection tours of the facility and meetings with licensee personnel to ascertain the status of the licensee's fire protection program and to verify conformance with applicable codes, standards and regulations.
The following plant areas were examined: control room; diesel generator and battery rooms; cable spreading areas; switchgear areas; safety related pump areas; intake structure; and, general areas in the auxiliary building and the containment. The above areas were examined for (as applicable): general conditions, including accessibility and congestion; cable tray and conduit separation; fire barrier and stops; fire detection and suppression equipment; ventilation systems and controls; floor drains; communi-cation equipment; lighting; coebustible materials; equipment pro-tection from water damage; means for :ontaining oil spills; and ability to contain fire, or isolate from external fire.
A detailed sur. aary of the review team findings will be presented in a separate report from the NRC. Within the scope of this inspection, the inspector had no further comments on this item at the presen ' .
3.
SI System Supports The licensee submitted LER 78-53 to NRC: Region I on October 27, 1978 to report errors identified by the licensee's A/E in the piping stress analysis completed for safety iniection piping inside of containment. The analysis errors reportedly stem from errors made in hand calculations for 6 inch diameter and smaller piping and affect the six hot / cold leg injection lines. The errors were discovered during an A/E review of the stress calculations after receipt of new information from the NSSS vendor which corrected the weights of check valves installed in the injection lines. The check valve weights were determined to be about 450 lbs. versus the 250 lbs. assumed in the original analyses. The stress levels were recalculated by the A/E and were determined to be in excess of the allowable stresses of ANSI B31.1, 1967.
The yield stress level was exceeded in one case out of six flow paths.
However, the A/E has concluded that no loss of safety function would have occorred under accident conditions even with the recomputed stress lovfis.
No additional information was available from the A/E regarding the analysis and its consequences as of November 3,1978.
The inspector stated that the NRC position was that the integrity of the SI system be demonstrated to confonn with the FSAR requirements prior to startup of the plant at the end of the current transformer outage.
The licensee acknowledged the inspector's comments. The inspector submitted to the licensee the following items which should be addressed during the licensee's review of subsequent information provided by the A/E: what segments of the SI piping were affected by the analysis -- errors? identify lines and locations of pipe supports; identify the hand calculations employed in the analysis; -- address how the hand calculations interface with other (computer) analysis; discuss the details of the hand calculation errors; what are the recalculated stress levels? what are the margins -- to the B31.1 allowable stress levels? to the yield stress levels? to thenultimate stress levels? -- provide a basis for the conclusion that there would have been no loss of safety function.
what assurance can be given to show that the calculational -- error applies only to the six points in question? to only the SI system? to only the BV facility?
. .
provide the details of any pipe support modifications that may -- be required? what will be the new stress levels and the margins to the -- allowable stresses? The licensee acknowledged the above information and stated that the 14-day followup report would address these items.
This item is unresolved pending subsequent review by NRC: Region I of the licensee actions in this area (50-334/78-30-01).
4.
Containment Floor Liner Welds During discussions with licensee personnel regarding the require-ments applicable to the Type A Integrated Leak Rate Test being conducted on BV Unit 1, the inspector noted that problems had developed during the fabrication of the containment floor liner on BV Unit 2.
The inspector interviewed licensee personnel and reviewed selected records to determine the nature of the problems experienced on Unit 2 and to ascertain whether the Unit 2 problems were appli-cable to the Unit 1 floor liner welds.
Licensee evaluation of the Unit 2 weld problems was documented in an internal A/E report entitled " Final Report on Reactor Contain-ment Liner Floor Plate to Bridging Bar Welds and Associated Backing Bar Butt Welds".
The problem was first recognized in 1977 when Unit 2 craft personnel noted flaws in welds where liner plate had been joined to heavier gauge bridging bar in the neutron shield tank, crane wall support and component support areas.
In most cases, the weld flaws were found to develop. as long as one week after welding had been completed, and subsequent to successful NDE on the subject welds.
Laboratory analysis of samples taken from the Unit 2 site indicated the flaws were attributable to hydrogen induced embittle-ment, as evidenced by the delayed nature of the cracks and from a study of flaw structure.
Other factors contributing to the cracking were determined to be: high weld stresses caused by high construc-tion stresses and the highly restrained configuration of the weld joint; poor welding performance due to inadequate drying of the weld area and application of heat treating; a d, the use of high strength weld materials which are more susceptible to hydrogen induced cracking. The licensee developed a program to correct the Unit 2 liner weld problem which has been successful._J1ocumentation of the corrective actions adopted will be found ii NRC:R'egion I RC&ES Branch ' ~ inspection reports for 1977 and 1978; resolution of the Unit 2 problem has been reviewed and found acceptable by NRC: Region. .
In response to this inspector's queries regarding the applicability of the Unit 2 problem to the work performed on the Unit 1 containment floor liner welds, the licensee stated that, based upon an evaluation completed in conjunction with the A/E, the Unit 2 weld problems were deemed to be unique to Unit 2 and were not applicable to Unit 1.
Some of the factors supporting this conclusion were cited to be: the smaller quantities and different geometry of the structural steel used in the design of Unit 1, which produced lesser construc-tion and configuration stresses than in the Unit 2 design; and, the weld materials used during Unit 1 construction.
The inspector noted that although this evaluation has been completed through a joint effort by the licensee and his A/E, the evaluation had not been formally documented.
The inspector stated that a formal documentation of the evaluation appeared to be appropriate.
The inspector stated that this item would be further reviewed by the NRC and is considered to be unresolved (50-334/78-30-02).
5.
Unresolved Items Unresolved items are those items for which more information is required to determine whether the items are acceptable or items of noncompliance.
Unresolved items are contained in paragraphs 3 and 4 of this report.
6.
Exit Interview A management meeting was held with licensee personnel (denoted in paragraph 1) at the conclusion of the inspection on November 3, 1978. The purpose, scope and findings of the inspection were dis-cussed as they appear in the details of this report. The licensee noted the inspector's concerns in regard to the SI system piping supports and the Unit 1 containment floor liner welds. }}