IR 05000454/1986042
| ML20214V084 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/01/1986 |
| From: | Danielson D, Jeffrey Jacobson, David Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214V034 | List: |
| References | |
| 50-454-86-42, 50-455-86-38, NUDOCS 8612090517 | |
| Download: ML20214V084 (6) | |
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U.S. NUCLEAR REGULATORY COMMISSION Y
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REGION III
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Reports No. 50-454/86042(DRS); 50-455/86038(DRS)
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g Docket Nos. 50-454; 50-455 Licenses No. NPF-37; NPF-60 i
Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2
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Inspection At: Byron Site, Byron, Illinois Inspection Conducted: October 14-17 and Novembe'r 25, 1986
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Date hhdb Approved By:
D. H. Danielson, Chief I L)' 2(,
Materials and Processes Date Section
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Inspection Summary-
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Inspection on October 14-17 and Novemberi25, 1986 (Reports No. 50-454/86042(DRS);
No. 53-455/86038(DRS))
Areas Inspected:
Routine, unannounced inspection of previous inspection findings, 50.55(e) itens, and Reactor Coolant System hydrostatic test results.
This report addressed the following Inspection Modules:
92700, 92701, 99014, 92702, and 70562.
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s1 Results: No' violations or deviations-were identified.
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DETAILS
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1.
Persons Contacted Commonwealth Edison Company (CECO)
- R. Klingler, Project QC Supervisor E. Martin, QA Superintendent W. Dijstelbergen, Field Engineer Nuclear Regulatory Commission (NRC)
J. Hinds, Jr., Senior Resident Inspector The inspector also contacted and interviewed other licensee and contractor employees.
- Denotes those individuals attending the final telephone exit interview on November 25, 1986.
2.
Licensee Action on Previous Inspection Findings a.
(Closed) Open Item (455/83000-23): Hydraulic Operators for Steam Generator PORVs SER Section 7.4.2.3 - Supplement 2.
The licensee committed to providing hydraulic operators for the steam generator power operated relief valves (porvs). This modification involved removing the air operator and replacing it with a hydraulic operator. The NRC inspector reviewed the completed work packages for both Units 1 and 2, and observed the final installation on Unit 2.
b.
(Closed) Open Item (455/85047-03):
IDI finding, Paragraph 5.3.c heat dissipation calculation used an assumed value of 95% of motor efficiency instead of actual vendor supplied value.
The Architect - Engineer (A-E) evaluated the open item and responded with an explanation of the method used for heat dissipation calculations. An assumed value of 95% motor efficiency is used for all 4 kv and 6.6 kv motors. This assumption simplifies the heat dissipation calculations. The results of the calculation are not affected by use of actual motor efficiencies for each motor.
c.
(Closed) Open Item (454/85011-04): Welder items of concern.
This item was closed in NRC Report No. 50-454/86017(DRS) as Open Item No. 454/85001-04.
d.
(Closed) Open Item (454/85050-01):
CAT Inspection items, bolting, wiring of valves and RT interpretation.
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l The licensee completed reviews and evaluations for Units 1 and 2, l
including generic implications as_a result of a Construction Appraisal Team-(CAT) inspection (Report No. 50-455/85-27) at the Byron Unit 2 facility.. The. licensee noted missing radiographs during the review.
Subsequent discussion between NRC Region III and NRR concluded that
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the missing radiographs.had no safety implications and was deemed-acceptable.
e.
(Closed) Open Item (455/86015-01): UT indications in steam-generators and pressurizer.
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The licensee concluded that the indications found in the steam generators and pressurizer-were either small slag inclusions formed during vessel fabrication or weld inner diameter surface geometry
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not cracks or lack of fusion.
The licensee proposed not to remove the indications based upon fracture mechanics analyses that indicate that the indications will not grow to an unacceptable size during plant life.
The licensee has agreed to perform inservice examination on the areas containing the flaws and secondary side hydrotest and leak
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tests.
f.
(Closed) Open Item (454/86031-10; 455/86017-10): Non-safety related
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attachment welds to safety-related structures were not Q.C.
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inspected as required by S&L Specifica1 ton F-2790.
In addition, transverse welds on the flange of beams, prohibited by S&L Specification F-2790 have been performed. The licensee has taken
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l the position that Q.C. inspection of these non safety-related welds was never intended. These welds are used to support plant lighting and communication conduit and fixtures.
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The issue of weld quality of these non safety-related welds lies
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I with the potential effect of any discrepancies in such welds on the structural integrity of the safety-related structural steel. The two types of weld discrepancies which can potentially affect the structural steel include planar defects (cracks and lack of fusion)
and volumetric defects (undercut).
To provide assurance that no design significant defects have
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resulted from this lack of QC inspection, the licensee performed ;
sampling inspection of the subject welds at both the Byron and Braidwood sites.
Sixty weldments were selected from each site which corresponds to the random sample size required for a 95/95
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confidence / reliability.
The welds were inspected using the recently adopted Visual Weld Acceptance Criteria. As a result of this inspection, over 90% of the welds satisfied the inspection criteria.
Engineering i
evaluations were performed on those welds not accepted by the
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inspection and it was shown that no discrepancy was large enough to be of design significance.
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The NRC inspector reviewed S&L Drawing No. 20 E-0-3000L, Sheet 1, Note 3G. This drawing pertains to conduit installation and now allows transverse welds on beam flanges for this application.
Cross flange welding for this application has been evaluated and accepted by S&L.
g.
(Closed) Violation (455/85027-01C):
Inconsistencies regarding the effective embedded length (Le) of concrete expansion anchors (CEA).
After installation of the CEA, it is not possible to precisely determine the location of the expansion wedges due to the movement of the anchor necessary to expand the wedges.
Since the qualification tests, installation specification, and contractor installation procedures all define Le as the distance from the concrete surface to the bottom of the wedges, a precise post-installation verification of Le is not possible.
The inspection method used by Pittsburgh Testing Laboratory (PTL) to verify minimum Le requirements assumes that the wedges are located at the bottom of the anchor after installation. This assumption is not conservative in that theoretically the wedges may be located as much as one anchor diameter from the bottom of the anchor after installation. An insufficient Le can reduce the concrete pullout capacity and is therefore of concern.
To resolve this concern, S&L has determined that the concrete shear core failure capacity, using an Le of one enchor diameter less than the required 8 diameters, is still greater than the anchor capacity determined by qualification for 3/8 inch diameter and larger anchors. Thus, the ultimate capacity of the anchors is not affected.
For inch diameter anchors, a sample of 60 assemblies including 612 individual anchors was chosen for evaluation. Of the 612 anchors, it was determined that 47 anchors could theoretically have insufficient Le when no anchor movement during installation is assumed. This is a conservative assumption in that some anchor movement is required to set the anchor.
For these anchors, the actual loads were compared to the reduced concrete pullout capacity and in all cases, the factor of safety was in excess of four.
Based on the engineering evaluations summarized above, the continued inspection of CEA installations utilizing the PTL method does not reduce the ultimate anchor capacity to any design significant degree.
3.
Licensee Action On 10 CFR 50.55(e) Items (Closed) 50.55(e) Item (455/86003-EE): Cold Leg Safety Injection Pipe Vibration and Cracks.
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During ECCS Full Flow Testing for Byron Station, Unit 2, in February 1986, a 1 1/2 inch socket welded-elbow on A-loop high head cold leg injection cracked and began leaking. Metallurgical examination revealed micro-cracks which had propagated through the wall. Vibration testing showed high vibrational loads occurring at the elbow which
. exceeded code allowable. On A and D loops, the elbows were replaced with five-diameter bent pipe (5-D Bend) and valves were relocated to avoid a
" lumped mass" near the point of stress concentration. Testing with strain gauges confirmed the change was acceptable and was within code allowable. No action was necessary on the B and C Loops since they do not experience the high vibration loadings of A and D loops and fall within the revised acceptance criteria. The NRC inspector reviewed the corrective actions and associated documentation and found them to be acceptable.
(Closed) 50.55(e) Item (455/86004-EE): Ultrasonic Indications were found in S/G's and Pressurizer Welds.
This Item is addressed in NRC Open Item No. 455/86015-01, Paragraph 2e above.
4.
Reactor Coolant System Hydrostatic Test Results Evaluation The NRC inspector reviewed the procedures, test data and related documentation pertaining to the Reactor Coolant System hydro. Marked up piping diagrams were reviewed to assure that the test boundary included pressure vessels, piping, pumps and valves up to and including:
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The outermost containment isolation valve for piping penetrating the primary containment.
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The second of two valves normally closed during normal operation in piping that does not penetrate the containment.
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Reactor Coolant System safety and relief valves.
The minimum test pressure of 1.25 times the design pressure was held for the required 10 minutes before lowering to the inspection pressure.
Static head calculations were reviewed to assure correct compensation for pressure gage location and to preclude overpressurization of low points in the system.
Pressure gage and temperature indicator calibration records were reviewed to verify the test validity. The reactor coolant test temperature was verified to satisfy the nil ductility requirements of the coolant system.
Follow up partial hydro tests were also reviewed to the above requirements.
In summary, the Byren, Unit 2 hydro appeared to meet ASME code requirements and is considered acceptable.
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5.
Exit Interview The inspector met with site representatives (denoted in Persons Contacted paragraph) at the conclusion of the inspection. The inspector summarized the scope and findings of the inspection noted_in this report. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents / processes as proprietary.
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