ML20031H359
| ML20031H359 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 09/18/1981 |
| From: | Torres M GENERAL ELECTRIC CO., HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20031H319 | List: |
| References | |
| NUDOCS 8110270405 | |
| Download: ML20031H359 (3) | |
Text
..
Scptcmbar 18, 1981 1
UNITED STATES O? AMERICA NUCLEAR REGULATORY COM>JISSION 2
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3
4 In the Matter of S
S HOUSTON LIGHTING & POWER COMPANY S
Docket No. 50-466 5
S (Allens Creek Nuclear Generating S
6 Station, Unit 1)
S 7
DIRECT TESTIMONY OF MARTIN R. TORRES REGARDING:
0 (1)
TEXPIRG CONTENTION NO. 11 - FLOW INDUCED VIBRATION (2)
DOHERTY CONTENTION NO. 31 - FLOW INDUCED VIBRATION /LPRM Q.
Mr. Torres, have you reviewed your prior affidavit 10 on TexPivj Contention No. 11 and Doherty Contention No. 31, 11 which affidavit is attached hereto as Attachment MRT-l?
13 A.
Yes, I have.
13 Q.
Are the' statements contained therein still true and 14 correct?
1~4 A.
Yes, they are.
16 Q.
Mr. Torres, have you reviewed TexPirg's response to 17 Applicant's motion for sm$ mary disposition of TexPirg Contention 18 11?
19 A.
Yec, I have.
20 Q.
Would you please comment on its response?
21 A.
First, I would note that the studies which have been 22 done are not " paper studies."
The testing program described in my affidavit is far more detailed.
This test program fully 23 mpli s with Reg. Guide 1.20 "Comprehensis: Vibra. tion 24 1
_j O$0k4fs PDR
..____.._. _q l
Assessment Program for Reactor Internals during Pre-operation 2
and Initial Startup Testing".
This guide presents a compre-henrive vibration assessment program for use in verifying 3
the structural integrity of the reactor internals for flow 4
induc d vibration prior to commercial operation.
As I 5
stated in my affidavit, the Applicant has agreed to comply 6
I with Reg. Guide 1.20.
This Reg. Guide specificall; endorses a
pre-operational testing as a method of assuring the absence of flow induced vibration.
Obviously, test results from tile 9
prototype, preoperational and initial startup tests described 10 in my affidavit cannot be made available until the tests are 11 performed which will occur after Perry I or Allens Creek is 12 constructed.
13 Q.
Have your reviewed Mr. Doherty's response to 14 Applicant's motion for summary disposition of Doherty 15 Contention 31?
16 A.
Yes.
17 Q.
Would you please comment on his response?
18 A.
First, the intervenor's contention is technically 19 misorientated.
Incore instrument tubes such as Local Power 20 Range Monitors (LPRMs) are made of stainless steel and placed adjacent to fuel channel made of Zircaloy.
When the 21 LPRM vibrated in an earlier design (not the same as Allens 22 Creek), the LPRMs were not damaged in any way as the stainless 23 steel did not wear but the adjacent Zircaloy fuel channels 4
r
~
1 did wear.
The cause of the LPRM vibration was due to inter-2 fuel bundle coolant (water) flow injected by one inch (l")
3 holes in the core plate directly below the LPRM.
In earlier 4
BWR designs, these holes were not present and the LPRM did n t vibrate and fuel channels were not damaged.
In plants 5
with these holes, which had subsequent fuel channel wear, 6
the holes were plugged.
These plants with 1" holes plugged i
show no degradation of LPRM function and no channel wear g
after operation since 1974.
In the Allens Creek design, plugging the holes was not necessary as they were not in the design (i.e., never drilled).
Years of BWR experience shows 11 no fuel channel wear, no degradation of incore instrument 12 tubes in any BWR design with no 1" holes drilled into the 13 core plate for interchannel flow.
14 Second, the intervenor's request for additional 15 LFRMs is totally meaningless.
The problem of flow induced 16 vibration on the LPRMs has been eliminated, and in any event 17 there was never any damage to the LPRMs from flow induced 18 vibration.
19 20 21 22 23 24
~
Attachment MRT-1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY
)
)
)
Docket No. 50-466
)
)
(Allens Creek Nuclear Generating
)
Station, Unit No. 1)
)
AFFIDAVIT OF MARTIN R. TORRES State of California County of Santa Clara I, Martin R. Torres, E iger, Flow Induced Vibrations, Nuclear Power Systems Engineering Department of the General Electric Company, of si lawful age, being first duly sworn, upon my oath certify that the y
statements contained in the attached pages and accompanying exhibits are true and correct to the best of my knowledge and belief.
Executed as San Jose, California p
July 29, 1980.
Subscribed and sworn to before me this1 day of July, O.
m
.M[oYA w
NOTARY PUBLIC IN AND FOR SAID COUNTY AND STATE h
Mv commission expires %M c2 /
of iggj
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RUTHE M. K!NNAMON OFFICIAL SEAL r
NOTARY PUBUC - CAUTORNIA
$At4TA CLARA COUNTY My comm. espires MAR 28.1981 i
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A,... sN;n.. CA 9si 25 V 'tN,,
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4O JDH:sem/1031
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1-7/29/80
Attachment MRT-1 f
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATC!!IC SAFETY AND LICENSING BOARD In the Matter of S
S HOUSTON LIGHTING & POWER S
COMPANY S
4 S
Docket No. 50-466 (Allens Creek Nuclear.
S Generating Station, Unit S
No. 1)
S a
- b Affidavit of Martin R.
Torres My name is Martin R. Torres.
I am employed by the General Electric Company as Manager, Flow-Induced Vibration, y
Nuclear Systems Engineering Department.
I have served in I
this capacity for five years.
A statement of my experience and qualifications is set out in Attachment 1.
This affidavit addresses TexPirg Contention 11 and
[.j ~
Doherty Contention 31.
These Contentions state that flow-
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induced vibrations on the following components have not been adequately assessed.for ACNGS:
(a)
Jet Pumps (b)
Spargers (c)
Fuel pins I'
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(d)
Fuel rods 9
(e)
Incore instrumentation g
(f)
LPRMs 6 sui.
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The phenomenon of flow-induced vibrations has been studied extensively on previous General Electric plants.
Information gathered from these studies and test programs 4
designed specifically to study flow-induced vibration has been
^
used to improve design and to qualify ACNGS.
Four sets of g
analyses and tests verify that flow-induced vibration will not impair the safety of ACNGS.
These are:
1.
A dynamic system analysis, n,
2.
Flow tests, forced oscillation tests, and other i
physical tests of reactor internal ccmponents.
3.
Prototype plant pre-operational and operational tests.
4.
Pre-operational tests of ACNGS.
Gj The dynamic system analysis is described in Section 3.9.1.3 of GESSAR 238 NSSS and has been in use since the kf,
licensing of Browns Ferry Unit 1.
This analysis described flow-induced vibration which may result from normal reactor operation.
This analysis serves two functions.
GE uses it during the design and in-house testing phase of reactor internal components.
The dynamic system analysis is also L.
used to establish criteria for plant pre-operational vibration r
testing (Item 3 above).
For example, such an analysis has
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been prepared for Perry Station Unit 1, the prototype 238 ll BWR-6 plant.
This analysis is fully applicable to ACNGS.
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Flow tests of various reactor internals to quantify flow-induced vibration levels are an integral part of General Electric's design process.
These tests are conducted to verify design and are independent of vibration testing required by NRC regulations.
The tests were performed at
.various test facilities starting as early as 1974 for BWR-6.
The tests-followed procedures required by 10 CFR 50, Appendix B.
In most instances, these tests were performed using full scale, actual reactor hardware at flow rates well in excess of the operational design condition.
Tests included both ay flow tests and, as appropriate, forced escillation tests.
Flow tests were performed on jet pumps, control rod guide
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low-pressure coolant infeculon lines, feedwater
- tubes,
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spargers, fuel assembly, in-core instrument tubes, and differ-ential pressure lines and other components, f
ad In one test facility, the High Flow Hydraulic Facility located at General Electric's Nuclear Energy Division in San Jose, California, a full scale mock-up of a segment of the BWR-6 core and lower plenum was flow tested to verify that flow-induced vibration amplitudes were within acceptable levels.
Other components, including in-core tubes containing instrumentation such as the LPRMs', fuel bundles and feedwater spargers were tested for flow-induced vibration in various other test facilities.
As an example, the feedwater sparger was flow tested at GE's Feedwater Sparger Test Facility in O
e
San Jose.
Fuel and in-core instrument tubes were flow tested in-the Building G, Large Tank Hydraulic Flow Loop in San Jose.
At the Pacific Gas and Electric Company's fossil power plant at Moss Landing, such components as jet pumps a
i.
were tested with steam and water at BWR operating conditions in a General Electric test facility.
Additional testing of G
arious components was done at the Colorado State University Hydraulic Laboratory, Fort Collins, Colorado, (one-fourth scale model of BWR-6) and the Atlas Test Facility, San Jose, California.
The vibration testing requirement of Regulatory 5
1/
L-Guide 1.20 will be satisfied on a prototype plant, presently designated as Perry Unit 1, whose operation is expected to precede that of ACNGS.
On the prototype plant, extensive f
vibration measurements will b6 made on major internal compo-nents, including the jet pumps, during pre-operational and p.t J
start-up flow testing, and an extended pre-operational flow test and inspection will detect evidence of possible undesir-ar 2/
able effects due to vibration.-
Vibratory responses will 0
be recorded at various recirculation flow rates and power levels, using strain gauges, accelerometeto and linear differential transducers, as appropriate.
Actual results of the data analysis, natural frequencies and mode shapes will then be compared to those obtained from the theoretical dynamic systems analysis discussed above.
Vibratory amplitudes will be compared to the criteria derived from that analysis.
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For the prototype plant, sensors will monitor possible flow-induced vibration during power operation for the jet pumps, spargers, LPCI coupling, core support structure, fuel channels, LPRMs and other in-core instrumentation.
From the fuel channel sensor measurements, information on E
possible flow-induced vibration effects for the fuel pins u
and control blades can be derived.
These tests will continue until power operating conditions are reached, and are scheduled to be completed prior to operation of ACNGS.
In the unlikely event 'h t ACNGS station becomes the prototype plant, the E
Applicant will follow Regulatory Guide 1.20 and perform 2/
these extensive pre-operational and operational tests.
Becauso ACNGS is not expected to ha the prototype q"
O plant for the vibration testing requirement of Regulatory Guide 1.20, confirmatory pre-operational flow-induced vibra-k
~
tion testing of reactor internals at ACNGS will be performed I
in accordance with the "non-prototype" testing provisions of 4/
The confirmation will be made by
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the use of extended high-flow testing, preceded and followed by a full inspection of internals in accordance with Regulatory o
Guide 1.20.
The extensive four-step testing and analysis
,ip.
program described above is fully e:tpected to eliminate
- a 5-flow-induced vibration at ACNGS.
Apart from this program,.
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however, at least four other factors provide assurance and protection against flow-induced vibration at Allens Creek:
1.
Reactor instrumentation can detect some vibration problems long before they poue any hazard.
2.
Design improvements will reduce the possibility of vibration damage to components such as the feedwater sparger, y
3.
ACNGS will have a loose parts monitoring system.
4.
Degradation or failure of some nonsafety components will not prevent a safe shutdown.
Experience has shown that reactor instrumentation 7f can detect some vibration problems long before they pose any hazard.
For example, instrumentation such as that measuring differe'ntial pressure, jet pump drive flows and pressures, feedwater flows and pressures and neutron sensors have, at U
^
tirues, shown anomolous performance that indicated possible vibration.
{
At Duane Arnold and Cooper nuclear plants, vibration in LFRM tubes was detected by examination of the transversing in-core probes (TIPS), based on the neutron ncire level in 5,6/
unfiltered TIP tracts.
This instrumentation also led to detection of in-core vibration at Browns Ferry.
The ACNGS design, however, eliminates the source of this vibratory 6-
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T-wear, which was traced to by-pass flow holes in the design of those plants.
By plugging these holes and prcviding an alternate flow path, the problem wTs eliminated.
Bypass flow holes are not a part of the design at ACNGS, so the same problem cannot occur.
Design improvements have produced components less likely to be damaged as a result of flow-induced vibration.
i:
For example, an improved interference fit feedwater sparger design will be employed at ACNGS.
The design consists of
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three concentric thermal sleeves, ensuring that detrimental E~
vibration is eliminated under all conditions.
ACNGS will have a loose parts monitoring system, an acoustic system designed specifically to detect any loose 7/
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parts in the reactor.
Degradation or failure of some nonsafety components such as feedwater spargers will not affect the capability of the plant to achieve and maintain a safe shutdown condition.
l In any event, should a feedwater sparger become inoperative, 3
a redistribution of inlet water temperature, flow rate, or in-core neutron flux would be detected by existing in-core instrumentation, allowing corrective action to be taken.
[
Ir conclusion, the lesson of nuclear plant operating history is that neither a loss of plant safety nor an inability L:
to safety shutdown the plant has ever occurred because of 1
L7 flow-induced vibration.
Moreover, all modifications to E
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a operating plants which have proven satisfactory in preventing flow-induced vibration have been rigorously implemented in the BWR-6 design.
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References 1/
Regulatory Guide 1.20, " Comprehensive vibration Asressment Program for Reactor Internals During Pre-operational and Initial Startup Testing," Rev. 2, May 1976.
2/
- Letter, G. G. Sherwood (GE) to E. G. Case (NRC),
Reactor Internals Vibration Assurance Program," MFN/169/78, April 7'
24, 1978.
3/
PSAR, Appendix C, p. C1.20-1.
4/
GESSAR-238 NSSS, Section 4.2.2.4.
5/
Lettera K. Goller (NRC) to D. G. Eisenhut (NRC),
Modifi-G cation to Eliminate Significant Incore vibration," dated March 2, 1976.
k 6/
Safety Evaluation Report by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, in the Matter of BWR Channel Box Wear, July 22, 1975.
7/
PSAR, Section 1.5.1.2.1.
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ATTACHMENT 1 m
Martin R. Torres 6
EDUCATION:
BSME,1965, University of Utah GE Advanced Engineo ing Co use Graduate, 19C5-1968 r
MSME,1968, Universr.y of Califernia, Berkeley
]
- , Applied Mechanics, Stanford University
~
g2:RIENCE:
Present Position:
Manager, Flow Induced Vibrations, Nuclear Power Systems Engineering Department, General Electric Company 1975-Present:
Manager, Flow Induced Vibrations
~
Responsibilities include management of engineers and technicians engaged in highly technical and experimental work in the field of flow induced vibration.
Mr. Torres is responsible for the flow induced vJration testing of all BWR reactor internals.
The world's g
largest test facility - the High Flow Hydraulic Facility - is under l
his active direction.
He was task force leader in several key FIV test programs, such as the feedwater sparger and incore instrument tube.
He carries management responsibility for the extensive FIV Program for Light ?iater Reacters funded by the U.S. Department of Energy frem 1976 to the present.
Reporting to him on this program are the FIV technical teams of Argonne National Laboratory and General Electric's Corporate Research and Development Center as well
!J ~~
as the NPSED Program Manager, Dr. Mark A. DeCoster.
In summary,
((
Mr. Torres has performed or directed every General Electric FIV experiment relative to the BWR since 1972.
J' 1972-1975:
Senior Developaent Engineer Primarily responsible for BWR flow induced vibration (FIV) test pro-a l
grams.
Plan, coordinate design hardware / instrumentation, data M
acquisition and analysis for FIV test program.
Extersive internal I:,
reports written in the area of flow induced vibration development testing:
1.
Feedwater Sparger Vibration Testing li.
2.
Jet Pump Vibration Testing i"
3.
Incore Instrument Tube Fuel Channel Testing I
4.
Scaling and Model FIV Testing 5.
FIV of Cylindreical Rods in Parallel Flow in 6.
FIV of Inclinad Cylindrical Rods in Longitudinal Flow 2
1968-1972: Dynamic Analysis Engineer - BWRSD 2
Seismic and vibration analysis of nuclear power plant equipment.
Brief list of analytical work performed:
.~
e Extensive graduate work undertaken from 1969-1979 at Stanford, Flufd Mechanics, Dynamics, Material Science, Solid Mechanics and Mathematics
(%100 graduate quarter hours).
T-
P-
'l.
Equivalent Damping Undcr Random Vibration 2.
Recire Icop Vibratic 4 Analysis 3.
Dynamic Loads Due t> LOCA 4.
Seismic Analysis of RPV and Internals 5.
Nonlinear Analysis of CRD Housings 6.
FIV Analysis of BFTR Componenets 7.
Prebabilistic Approach to Seismic Analysis 8.
Fatigue Useage of BWR Internals Several papers published in ASME Transactions and earthquake engineering literature on above subjects.
1-1965-1967:
NED Engineering Rotation Program Themal-Hydraulics - Transient Analysis B'n3 Thermal-Hydraulics - Fast Steam Cooled Reactor Stress Analysis ny!neer - Fast Flux Test Facility Product Design Engineer - Nuclear Instrumentation Department Since 1969, Mr. Torres has been a lecturer in dynamics and mechanical vibrations for General Electric's (internal) Advance Engineering Program.
Mr. Torres is a registered California Professional Engineer (Lic. No. 14674), a Member of ASME and the following Honor Societies:
Pi Tau Sigma Tau Beta Pi Phi Kappa Phi r
Magna Cum Laude, 1965 BSME E
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