IR 05000259/1993011
| ML18036B264 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/27/1993 |
| From: | Blake J, Chou R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18036B263 | List: |
| References | |
| 50-259-93-11, 50-260-93-11, 50-296-93-11, NUDOCS 9305040259 | |
| Download: ML18036B264 (11) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/93-11, 50-260/93-11,= and 50-296/93-11 Licensee:
Tennessee Valley Authority 3B Lookout Place 1101 Harket Street Chattanooga, TN 37402-2801 Docket,Nos.:
50-259, 50-260 and 50-296 Facility Name:
Br owns Ferry 1, 2, and
Inspection Conducted:
Harch 29 - April 2, 1993 License Nos.:
DPR-33, DPR-52, and DPR-68 Inspector:
R ch C.
Chou
~~7-'t 3 Date Signe Approved by:
.
Bl ake, Chi ef H
erials and Processes Section ngineering Branch Division of Reactor Safety
~7 P'te Signed SUHHARY Scope:
This routine, announced inspection was conducted in the areas of Inservice'nspection (ISI) and Reactor Vessel Stress Calculations.
Results:
In the areas inspected, violations or deviations were not identified.
Two unresolved items were identified.
One involved a case where the weld shapes assumed in the support calculations appeared to be different than the actual weld shapes fabricated in field, (paragraph 2.b.)
Another involved the explanation required to clarify the stress intensity shown in the summary stress report for Reactor Vessels 1 and 2, (paragraph 3.)
'P305040259 93042'V PDR ADOCK 05000259
REPORT DETAILS Persons Contacted Licensee Employees R. Baird, Civil Engineer
- M. Bajestani, Technical support Manager
- J. Corey, Radcon Manager
- C. Crane, Maintenance Manager
- J. Haddox, Engineering Manager
- H. Herrell, Operation Manager
- J. Johnson, guality Assurance Hanager
- L. Madison, Recovery Civil Engineer
- D. Massey, Licensing Engineer
- G. Pierce, Site Licensing Manager
- E. Ridgell, Licensing Manager
- J. Rupert, Engineer and Maintenance Manager
- 0. Zeringue, Site Vice President Other licensee employees contacted during this inspection included craftsmen, mechanics, technicians, and administrative personnel.
NRC Resident Inspectors
- C. Patterson, Senior Resident Inspector
- J. Hunday, Resident Inspector
- R. Musser, Resident Inspector
- G. Schnebli, Resident Inspector,'T.
Liu, Resident Inspector (Intern)
- Attended exit interview Preser vice and Inservice Inspection (PSI 8 ISI) - Unit 2 This refueling outage for Unit 2 is the first refueling outage since the plant was restarted in 1990.
a ~
Status of ISI At the time of the inspection, the licensee had almost completed all the 'scheduled ISI NDE for this refueling outage.
Left to do were eight UT examinations on pipe welds and some PSI weld examinations on replacement piping.
The table below summarizes the ISI and PSI NDE activities.
The only rejectable indication to this date was a cracked nut which was scheduled to be replace TOTAL LOCATION
680 182 Washers and nuts on Recirculation Pumps and Reactor Vessel Pipe welds on stabilizers, RPV welds, and Turbine valve guides, seats, and bolts Pipe welds on stabilizers, RPV welds, RPV head studs and nuts, turbine rotors and blades
128 2266 Various components EECW system pipe welds, RWCU pipe replacement, wetwell vent 'piping Pipe welds, RPV head studs and nuts, Recirculation Pump Studs, Turbine bolts and valves, and RWCU pipe replacement Field Observation of PSI on RWCU Pipe Replacements.
The reactor water cleanup (RWCU) piping system was replaced during this refueling outage.
The line included about 300 feet of piping and 30 pipe supports.
All of the supports were removed and reinstalled due to the pipe replacement.
Only support No. 2-47B406-S0018 and the anchor support at penetration X14 were required to be preservice inspected.
The inspector used the support drawing, and the licensee's visual inspection procedure, to re-inspect support No. 2-47B406-S0018.
This support had been inspected by the licensee's ISI group and determined to be acceptable.
The procedure used was NO. N-VT-1,
"Preservice and Inservice Visual Examination Procedure",
Rev.
18.
The inspector observed welding 'activities on the anchor support at Penetration X-14.
The support was about 95 percent reinstalled, with the exception of the welding of some stiffener plates to reinforce the connections between the members and the embedded plates.
The observation included welding, as well as guality Control (gC) fit-up inspection and review of the finished welds and erected members.
During the review of finished welds and comparison to the design drawings, the inspector noted that the welds at two connections between a horizontal wide flange beam and a vertical wide flange post had discontinuous'weld segments at one side of the wide flange web.
The three weld segments consist of a discontinued U-shape which had no welds at the two corner The inspector questioned the field engineer, who worked for the contractor General Electric Company (GE), why the weld was made as a discontinued U-shape without welds at the corners.
His response was that the drawing only indicated weld on three sides (three segments)
at one side of the wide flange web; therefore, it was unnecessary to make a continuous U-shape weld which included the corner welds.
In addition, the field engineer showed the inspector another sketch, with a corner weld to be added, to demonstrate that his understanding was that if the corner welds were required, the drawing would indicate them.
The inspector discussed the installed U-shaped weld configuration with. the licensee's engineers and requested a copy of the
=calculation for this support to see which shape of weld was used to qualify the connections.
The members and weld connections for this support were qualified by using Bechtel computer program
, FAPPS HE150.
The licensee's engineers showed the inspector page 5-7, Figure 5. 1, of the user's manual for FAPPS, which includes five structural shapes (or member shapes)
and 21 weld types (weld shapes or configurations).
Host weld types or configurations require a continuous weld at the corners.
The exceptions were some tube steel connections which do not require welds at the corners.
After reviewing Figure 5. 1, the inspector told the licensee's engineers that unless corner welds are specified in the support drawings as special notes at particular welds or in the general notes of the drawings, the welders are not obligated to make corner welds.
Therefore, the design engineers should either specify the special requirements for corner welds, in the drawings, or take a conservative approach by not using the corner welds in the computer programs or design calculations.
During discussions with the inspector, GE welding supervisors tried to explain that the welder and field engineer had misunderstood the weld symbols.
The supervisor showed the inspector that Page 3, of Welding Procedure G-29 lists American Welding Society (AWS) A2.4 as a reference, and ANSI/AWS A2.4-86, Figure 9 - Designation of Extent-of Welding, states that the
"desired weld" for a U-shaped weld connection is a continuous weld, which includes the corner welds.
The inspector disagreed with the welding supervisors understanding of the problem for the following reasons:
Figure 9 of AWS A2.4 shows a continuous U-shape weld, with corner welds, as a "desired weld" even when the weld symbol shows three segment welds.
A "desired weld" is not a
mandatory requirement for the weld, just a preferred wel AWS A2.4 is listed as one of six references on that page of the welding procedure.
The purpose of a welding procedure should be to provide necessary ins'truction to welders, without the need for them to read all referenced materials.
The welders are required to follow design drawings and welding procedures.
Based on the observation of field welding; review of procedures and drawings; and discussion with the design engineers, field engineers, and welding supervisors,,the inspector believes that there is a communication problem between the design engineers and the welders.
The licensee should therefore review, and revise as necessary Figure 5. 1 of Users Hanual of FAPPS HE 150; support drawing general notes; Welding Procedure G-29; and other pertinent documents to provide adequate communication so that welders will fabricate the weld connections that the design engineers want and assume in the design calculations.
Pending on the licensee resolution of this problem, this item is identified as Unresolved Item 50-259, 260, 296/93-11-01, Weld Differences between the Welds Assumed in the Design Calculations and Actual Welds Provided in Field.
No Violations or Deviations were identified in this area.
Stress Calculation Review - Units 1 and
The inspector partially reviewed Volumes
and 2 of "Summary Stress Report for General Electric - NED, TVA ¹I and ¹2 Reactor".
Babcock
&
Wilcox (B&W) Company was a subcontractor to General Electric (GE) to prepare this summary stress report.
This report was certified by a B&W Professional Engineer to the 1965 Edition of the ASHE Section III Boiler
& Pressure Vessel Code for Nuclear Vessels, including addenda through Summer 1965 and applicable Code cases.
This summary stress report consisted of seven volumes to, include a-summary of results and
reports (or sub-reports).
All reports were based on various GE computer programs.
GE maintained the computer input and output; the licensee received the seven volume summary stress report.
Volume 1 contained the summary report for the results from the
reports and reports No.
1 through 3.
The inspector reviewed the summary report for the results of each design report for each component of the reactor and the variations from the design and impact to the structural integrity of the reactor vessel for Units 1 and 2.
The stress allowables for the various materials and the detail dimensions for each
'omponent were also included in the summary of the report.
Volume 2 contained Report No.
4 for the stress analysis of the feedwater nozzle.
The report included results, conclusion and five transients.-
The report covered the stress analysis of the feedwater nozzle for mechanical load, steady state, operating, and transient condition This report demonstrated the evaluation of the primary stress intensity, the primary plus secondary stress intensity, and the peak stress intensity for the -feedwater nozzle, per the ASME Code requirements.
The primary stress intensity was evaluated for compliance with the 1.5 S
(S is the design stress intensity value) criteria of paragraph N-414 of ASHE 1965 Code Edition.
The primary plus secondary stress intensity was evaluated for compliance with the
S criteria of the same paragraph N-414.
The peak stress intensity w'as evaluated to-assure that the cumulative usage factor did not exceed the limit of 1.0 as permitted by paragraph N-415. If a component could not be qualified with the primary plus, secondary stress intensity to the
S criteria, the component can be qualified by using the peak stress intensity against the cumulative usage factor per paragraph N-415.
The peak stress intensity is defined as the highest value at any point across the thickness of a section of the combination of all primary, secondary, and peak stress produced by specified service pressures and other mechanical loads, and by general and load thermal effects associated with normal service conditions and including the effects of gross and local structural discontinuities.
Therefore, the value of peak stress intensity should be greater or equal to the combination of primary plus secondary stress intensity.
During the review of the stress summaries and calculations, the inspector noted some inconsistencies.
These inconsistencies are summarized as follows:
On page A-15 of Volume 1, the primary-plus-secondary stress intensities shown on the sketch exceeded the allowable stress intensities listed on the bottom of the same page.
Justification for the overstress was not provided.
On pages 8-16-1 to 8-16-16 of Report No. 4, in Volume 2, numerous figures or values
'shown in the tables for the peak stress intensity were lower than those shown in tables for the primary-plus-secondary stress intensity for the same juncture and transient.
(For example:
For the maximum stress at 'juncture 9 and transient No.
5 for L-H inside condition, on page P.B-16-2, 11 Ksi is shown for the primary-plus-secondary stress intensity and on page P.B-16-11, 0 Ksi is shown for the peak stress 'intensity.)
The fatigue analysis on pages 8-17-1 to 6, of Report 4, may not be valid because the peak stress intensity is not the highest value.
The final maximum primary plus secondary stress intensities shown on page 8-16-10 of Report 4 were not traceable.
Pending the licensee resolution to these questions, this item is identified as, Unresolved Item 50-259, 260/93-11-02, Explanation of-the Stress Intensity shown in the Summary Stress Report for Reactor Vessels 1 and No violations or deviations were identified in this area inspected.
Status of Unit 3 The inspector did not perform any inspections on Unit 3 activities"and only requested that the licensee provide a status report for information and inspection planning purposes.
The active items are listed below:
PROGRAM NO.
DCNs EST IHATED ESTIMATED ESTIMATED NO.
SUPPORTS/
ENGINEERING CONSTRUCTION COMMODITIES COMPLETION COMPLETION TOTAL TO GO TOTAL TO GO IE BULLETIN 79-02/14 LTTIP TORUS ATTACHED CRDHt SHALL BORE MVAC UPPER DRYWELL PLATFORHS LOWER DRYWELL PLATFORHS
18
,0 1800 500,
800
400 175 400 11/29/93 06/01/94 06/30/93 06/01/94 06/01/94 11/15/93 06/01/94 02/01/94 07/01/93 10/Ol/93 MISC. STEEL SUPPORT FRAMES
MISC.
STEEL PLATFORMS TORUS STRUCTURE
CABLE TRAY SUPPORTS*
CONDUIT SUPPORTS*
225
23
225
23
11-15/93 06/01/94 08/16/93 02/01/94 10/04/93 06/01/94 09/01/93
02/01/94
30
11/01/93 06/01/94
- SCOPE IS TENTATIVE.
SCOPE WILL BE DETERHINED BY FIELD NEEDS,
Exit Interview The inspection scope and results. were summarized on April 2, 1993, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
(Open)
Unresolved Item 50-259, 260, 296/93-11-01, Weld Differences Between the Welds Assumed in the Design Calculations and Actual Welds Provided in Field - Paragraph 2.B (Open)
Unresolved Item 50-259, 260/93-11-02, Explanation of the Stress Intensity Shown in the Summary Stress Report for Reactor Vessels 1 and
- Paragraph 3.