ML20032C699
| ML20032C699 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/06/1981 |
| From: | Beeman G Battelle Memorial Institute, PACIFIC NORTHWEST NATION, Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17198A148 | List: |
| References | |
| NUDOCS 8111100753 | |
| Download: ML20032C699 (8) | |
Text
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T NOV 0 61981 o
UNITED STATES OF AMERICA fiUCLEAR REGULATORY COMMISSION f
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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UNION ELECTRIC COMPANY Docket Nos. STri 50-483
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STN 50-486 (Callaway Plant, Ur~.ts 1 and 2)
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TESTIMONY OF G0RDON BEEMAN y.
Please state your name and qualifications?
A.
My name is Gordon Beeman.
I have been a research engineer for Battelle Memorial Institute at the Pacifk. :;orthwest Laboratory (PNL) since 1975.
I received a Bachelor of Science degree in Physical Metallurgy in 1972 and a Master of Science degree in Mechanical Engineering in 1975, both froa Washington State University.
I have been a ccr.hultant to the Mechanical Engineering Branch of NRC since August of 1979.
Since that time I have assisted in safety re/iews of seve. different nuclear plants currently under construction.
In t
addition.
I have directed the independent confirmatory stress analyses performed at Pill for the NRC to verify American Society of Mechanical Engineers (ASME) Code compliance for various safety related piping systems in ten different nuclear plants currently under construction.
Q.
What is the purpose of this testimony?
A.
The purpose of this testimony is to respond to Joint Intervenors' Contention II. A'.1 c.oncerning a piece of SA-358 piping.
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Q.
Please describe the history of your involvement with this matter?
A.
My invnivement in the matter raised in Contention II.A.1 came about as the result of an independent confirmatory piping stress analysis to (*;ertain compliance with the ASME Code. By coincidence the safety related system chosen for analysis at the Callaway plant Unit I was the loop No.1 emergency core cooling system attached to the reactor coolant loop inside containment. This includes the accumulator discharge line and part of the residual heat removal system. The results of tnis analysis are documented in a report submitted to the fiRC entitled "SNUPPS (Callaway and Wolf Creek) Loop No.1 Emergency Core Cooling System Piping Analysis," dated May 1981.
Q.
Could you describe the portion of the line of pipe involved in Joint Intervenors' Contention II. A.1?
A.
That portion of the line questioned in Contention II.A.1 is correctly designated as ASME Code Class 2.
It runs from the accumulator tank outlet nozzle to the check valve at the Class 1 to Class 2 boundary.
This portion of piping has a design pressure of 700 psi and a design temperature of 150 F.
Tne results of the aforementioned confirmatory piping stress analysis show a 7,000 psi stress at.he loci, ion of the alleged defective weld for sustained loads as calculated gr the equation in ASME Code NC-3652.1. The ASME Code allowable stress frr the piping naterial at 150 F is 18,300 psi.
Q.
Could you comment-on the allegation, in Contention II.A.1 that the pipe was substantially out-of-round?
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A.
In terms of this allegation, I have reviewed the measurements submitted by Applicant, Union Electric Company. The ASME Code contains no provisions for out-of-roundness of Class 2 cnd Class 3 piping.
However, the !!dterial Specification for ASME SA-358 piping allows an outside diameter variation of 1%. The pipe in question is 10 inch schedule 140.
It has a nominal outside diameter of 10.75 inches.
The measured outside diameter variation of the pipe at Callaway is 0.092 inches. This ! less than the 1% variation allowed by Material Specification.. *tE SA-358.
The ASME Code provides additional rules for out-of-roundness,^ Class 1 piping.
If the out-of-roundness is greater than 0.08t (wher is the nominal wall thickness), the stress index K1 used in NB-3653.i Jst be adjusted accordingly. The piping system in question is not part of the reactor coolant pressure boundary and is correctly designated as ASME Code Class 2.
Thus, this piping is not substantially out of round. However, even if the piping were evaluated using ASME Code Class 1 rules, the effect of the out-of-roundness would be negligible.
Q.
Could you comment on the allegation that the pipe was' machined below the minimum wall.
A.
I have independently performed calculations (using ASME Section III, Article NC-3640) to determine the acceptable minimum wall thickness for this pipe. These calculations reveal that a minimum wall thickness of 0.814 inches is acceptable.
Q.
Could you comment-on the allegation the* the pipe contained rejectable weld defects on the inside of a longitudinal seam weld?
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A.
I have not physically inspected the pipe, nor have I seen a radiograph of the weld.
I have been infonaed that radiographs have in fact been taken of the weld. Radiography is a method of nondestructive examination widely used throughout the industry to detect defects of this kind.
It will reveal any significant weld defects with a reasonable 4
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00!!D0411. BEDM, Senior Roscarch Engineer, Structures an'i Mechanics Sectio EngineeringPhysicsDepartment i
Education i
B.S.
Physical Metallurgy, Washington State Univerpity 1972 M.S.
Mcchanical Engineering, Washington State UniNersity 1975 Design Analysis In Long-Life Damage Tolerant!
Structures, UCLA extensica course 1977 Introduction to fracture Mechanics - knerican 1978 Society for Metals short course 1978 l
Dynamic Analysis Workshop, The Spectral Inst <itute i
5th Annual Conference on fracture Mechanics II, 1979 UnionCollege,Schenectady,NY Project Management, Battelle, Battelle Seminars 1980 and Studies Program Computer Workshop in Earthquake and Structuqa1 1980 Dynamics, Union College, Schenectady, NY Experience Mr. Beeman joined Battelle-Northwest in 1975, ininediatcli following completion of his graduate e rk.
He has worked on a wide variety of programs in the 5
structural analysis and material properties area.
Some c'f the programs in which he has been a major contributor include:
Mr. Beeman is currently the principal investigator for a team supporting a
I the Mechanical Engineering Branch of the Nuclear Regulatory. Commission.
The team is preparing safety evaluation reports in fulfillment of require-ments for nuclear reactor license applications.
Final safety analysis reports are reviewed for their compliarce with ASME Bditer and Pressure Vessel Code,Section III, ANSI B31.1,10cFR Part 50, NRC Regulatory Guides.
l Standard Review Plans and EB Technical Notes.
Part qf the effort in reviewing FSAR's involves computerized analyses of piping systems for their integrity under dynamic, thennal. pressure and deadweight loadings.
I Plenum Fill Experiment - Stress _Ana_1ysis Mr. Beesen assisted in the design stress analysis of the large simulator test vessel.
He was responsible for the analysis of (everal nozzle penetrations.
A viewport and flange were also analyzed using finite All components were designed accdrding to the ASME element techniques.
Boiler and Pressure Vessel Code,Section VIII Divisidns 1 and 2.
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e Extend Wire Rope Life o
During this project. Mr. Beeman assisted in the design of the wire rope testing facility.
He is currently project engineer in charge of fatigue testing and data evaluation for large diameter wire rope.
This project i
has dealt with evaluation of 3 in. diameter wire rope in bend-over-sheave fatigue under various loading conditions.
Special attention has been
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directed at correlating results from large-diameter rope with that from small-diameter rope.
Ropes from field appitcations are also being examined in order to verify test results.
Current wohk includes continued testing of large and small diameter metallic core wir( rope.
A remote 1
load sensing system is being developed to measure actsal loads on a wire rope during dragline operation at surface coal mines. :
Structural Evaluation of the Retrieva_ble_ Surface Storage Facility (RSSF) f Mr. Beeman worked on evaluation of candidate cask closure systems for various postulated accident conditions.
The study was funded by Atlantic Richfield Hanford Company for ERDA.
Additional work was done in the area of heat transfer and development for material prvperties for creep evalva-tion of the internal canister used for containment of solidified radioactive
- wasta, Dry Cooling Towe_r Mr. Beeman developed several cost algorithms that related dry cooling tower parameters to structural costs.
These algorithms were used as part of a large conputer program that is used as an aid in the design of dry cooling towers.
Additional effort was directed at obtaining data on the service life of polyethylene and polypropylene tubing to be used in the towers.
l structuraLAnalysis of_ Bubble Chaptber3m_ era Window __
o Thf s job was conducted for the Fermi National Accelerator Laboratory.
A structural and fatigue evaluation of a camera window for the high energy particle bubble chamber was completed.
Evaluation of the interaction between the glass window, an indium seal, and invar spring and a teflon seal was calculated.
The system was evaluated at ambient and liquid hydrogen temperaturts.
Evaivation of Mining Equ_ipment for the Z-9_Trinch_
Mr. Seeman acted as an outside reviewer for the structural calculations for the Z-9 mining facility.
Ass aspects of the design were reviewed as well es an onsite inspection of the factitty.
i Repnmissioning and Decontamination of Existing lianford Facilities Mr. Beemrn is currently engaged in developing a concrete testing program to deters.ine the service life of concrete used to isolate radioactive wastes from the environnent.
Special attention is being given to those pmporties of concrete necessary to provide the long term storage that is necessary.
Several methods for isolating radioactive waste with concreta are also being studied.
fuel Claddino Interaction Analysis e
Mr. Beesen is analyzing the interaction between the fuel pellet and the zirconium cladding in power reactor fuel rods.
A three-dinensional finite element computer assisted analysis is being perfonned.
An experit: ental apparatus is being designed so that testing can be done under controlled conditions that accurately simulate in-service conditions.
The prediction and prevention of fuel rod failures caused by fuel cladding interactions is the goal of this work.
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e Publicatien$
" Relative Evaluetion of Structural Integrity in Candidate Cask Closures for the Refere~nce Scaled Storage Cask Design", BNWL-B-442 Battelle Pacific Northwest Laboratories, October 1975.
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" Design Report for the Plenum fill Experiment Large Simulator Vessel", PfE 76107, Botte11e Pacific Northwest Laboratories, July 1976.
" Interim Report - Phase 11 Bend-Over-Sheave Fatigue Testing of Three-Inch Diaireter Wire Rope," $AM 77-2, August 1977.
"f actors Affecting the Service Life of Large Diameter Wire Rope," 2311102421, March 1978.
" Bend-Over-Sheave ratigue Testing of Large-Diameter Wire Ropes Used in Surface Mining" PNL-SA-7627, April 1979.
" Seismic Analysis Spent fuel Boil-Off Calorimeter and Support Structure",
SAM-79-1, April 1979.
" Wire Rope leprovement Program Fiscal Year 1981", PNL-3976, September 1981.
"Susquehanna Steam Electric Station Main Steam Safety Relief Yalve Line "N" Piping Analysis", SAH-80-11, K.I. Johnson, G.H. Beeman, September 1980.
" Polo Venie fluclear Generating Station Safety Injection System - Loop 1 A",
SAM-81-2, C A. Wf11iams, G.H. Beeman, May 1981.
"SNUPPS (Callaway and Wolf Creek) Loop No.1 Emergency Core Cooling System Piping Analysis", SAM-81-3, K! Johnson, G.H. Beeman, May 1981.
"Clinton No.1 Power Station Low Pressure Core Spray and Residual Heat Removal Piping Analysis", SAM-81-6, G.D. Marr, G.H. Beem6n, July 1981 "CE System 80 - Loop 1 Piping Analysis", SAM-81-9, C. A. Williams, G.H. Beeman, September 198i.
Professional Affiliatio_ns, Mr. Beeman is a member of the American Society for Metals.
He also has an Engineer in Training Certificate.
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