ML20055G574
| ML20055G574 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/19/1990 |
| From: | Wang A Office of Nuclear Reactor Regulation |
| To: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| IEB-88-002, IEB-88-2, NUDOCS 9007230332 | |
| Download: ML20055G574 (7) | |
Text
__..
July 19,1990 Docket No. 50-213 Mr. Edward J. Mroczka Senior Vice President Nuclear Engineering and Operations Connecticut Yankee Atomic Power Company Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 05141-0270
Dear Mr. Mroczka:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING BULLETIN 88-02, " RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM GENERATOR TUBES" The NRC staff has reviewed your analysis, provided in JACOR report J5439-89-001-R1, "Fluidelastic Instability Analysis of the U-bend Region of a Westinghouse Model 27 Steam Generator," in response to the subject bulletin and has determined that additional information is needed to complete our safety evaluation. Enclosed are our questions regarding-this topical report.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
If you have any questions or comments please contact me at (301) 492-3475.
Sincerely,
/s/
Alan Wang, Project Manager Project Directorate 1 4 Division of Reactor P ojects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated ccw/ enclosure:
See next page y
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DISTRIBUTION DocketlEile?
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Mr. Edward J. Mroczka c
Connecticut Yankee Atomic Power Company Haddam Ncck Plant cc:
Gerald Garfield, Esquire R. M. Kacich, Manager Day, Berry and Howard Generation Facilities Licensing Counselors at Law Northeast Utilities Service Company City Place Post Office Box 270 Hartford, Connecticut 06103-3499 Hartford, Connecticut 06141-0270 W. D. Romberg, Vice President D. O. Nordquist Nuclear Operations Director of Quality Services Northeast Utilities Service Company Northeast Utilities Service Company Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 s
Kevin McCarthy, Director Regional Administrator Radiation Control Unit Region I Department of Environmental Protection U. S. Nuclear Regulatory Commission State Office Building 475 Allendale Road Hartford, Connecticut 06106 King of Prussia, Pennsylvania 19406 Bradford S. Chase, Under Secretary Board of Selectmen Energy Division Town Hall Office of Policy and Management Haddam, Connecticut 06103 80 Washington Street Hartford, Connecticut 06106 J. T. Shedlosky, Resident Inspector Haddam Neck Plant E. A. DeBarba, Nuclear Station Director c/o U. S. Nuclear Regulatory Commission Haddam Neck Plant Post Office Box 116 Connecticut Yankee Atomic Power Company East Haddam Post Office RFD 1, Post Office Box '.27E East Haddam, Connecticut 06423 East Hampton, Connecticut 06424 G. H. Bouchard, Nuclear Unit Director Haddam Neck Plant Connecticut Yankee Atomic Power Company RFD 1, Post Office Box 127E East Hampton, Connecticut 06424
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l' ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION NORTHEAST UTILITIES e
HADDAM NECK i
LICENSE ~ ~~' '^NSE TO BULLETIN 88-02
References:
1.
Letter dated May 25, 1989, from Northeast Utilities responding to NRC Bulletin 88-02, 2.
JACOR report J5439-89-001-R1, "Fluidelastic Instability Analysis of the
' U-bend Region of a Westinghouse Model 27 Steam Generator," July 25, 1989.
Background:
Reference 1. reported that there are no tubes at any Haddam Neck steam generator which are subject to conditions which would lead to a North Anna type steam generator tube failure.
Reference 2 was provided to the NRC staff' in January 1990 and describes the technical basis for the conclusions in Reference 1.
The staff and its contractor, Argonne National Laboratory. have reviewed Reference 2 and request the following additional information.
Roquested Information 1.
.NRC Bulletin No. 88-02 (February 5, 1988) requests a program to assess the potential for steam generator tube failure in the U-bend region.
The program is outlined in subparagraphs (a) and (b) on pages 4 and 5 of bulletin and is based on an assessment re16tive to the failed North Anna tube.
Consistent with the bulletin request and as also recommended by JACOR (Reference 2, page 5-2), the model used to analyze the Haddem Neck steam generator tubes should be used to analyze the fluidelastic instability of tube R9C51 at North Anna 1 and the results provided to the NRC.
Tube R9C51 experienced a failure due to a fluidelastic instability and thus represents a benchmark for qualifying the JACOR model.
Relevant steam generator parameters (geometry, flow rates, circulation ratio, pressures, temperatures, etc.) for North Anna 1, existing prior to the July 1987 tube failure, can be obtained from either Virginia Electric Power Company (VEPCO) or the NRC staff.
2.
Why does ATH0S code give slip (unequal phase velocities) in axial direction only?
(Reference 2, page 3-1) 3.
Provide the detailed cal:ulations of K-loss factors which are documented on design analysis sheets made available to the licensee by JACOR.
(Reference 2, page 3-1).
?
4 2..
4.
JACOR notes that in the actual steam generator the tube support will likely represent something less than a completely fixed condition (page
~5-7).
To account for support flexibility JACOR uses translational and rotational stiffnesses identified in,s 1986 study by Schoof (Ref 6 of the JACORreport).
How did Schoof arrived at these values? Were they determined from tests on an actual steam generator?
5.
The basis for JACOR selecting a value of 0.03 for damping factor is apparently the work reported in Ref. 7 of the JACOR report; the staff does not have this reference. Other data that could be used for this application include that of Axisa et al (1984) [" Vibration of Tube Bundles Subjected to Air-Water and Steam-Water Crossflow:
Preliminary Results on Fluidelastic Instability". Symp, on Flow-Induced Vibration, vol 2, 269-284]. The results of Axisa et al give a damping value of approximately 0.009 (see Attachment 1) for a-void fraction of approximately 0.88.
This suggests that the value of 0.03 selected by Jaycor is nonconservative.
Please respond and include a copy of Ref.-7.
6.
On page 6-6, JACOR correctly notes that all three velocity components may lead to a crossflow condition.
However, in u,p, vgap, and wgap g
and their stability analysis they only use the transverse comporants u,p g
This assumes that flow in the plane of the U-bend v,p; see Eq. 6-10.
g does not affect fluidelastic instability. What is the basis for neglecting the velocity component in the plane of the U-bend? As illustrated on Fig. 4-3, this component is typ' eally the largest of the three.
I
SYMPOSIUM i
on FLOW-INDUCED VIBRATIONS J
I VOLUME 2 I
I VIBRATION OF ARRAYS OF CYLINDERS IN CROSS FLOW
-1 4
Presented at THE ASME WINTER ANNUAL MEETING I
NEW ORLEANS, LOUISl ANA DECEMBER 9-14,1984
\\
Symposium co sponsored by Applied Mechanics, Fluids Engineering, Heat Transfer, Noise Control and Acoustics, Nuclear Engineering, and Pressure Vessels and Piping Divisions i
Sessions in this Volume co-sponsored by s
l PRESSURE VESSELS AND PIPING AND NUCLEAR ENGINEERING DIVISIONS Edi:ed by M,P. PAIDOUSSIS (Principal Editor) g McGill University Montreal, Quebec, Canada i
M.K. AU YANG Babcock & Wilcox Lynchburg, Virginia S.S. CHEN Argor'ne National Laboratory Argonne, Illinois i
THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS l
j 1,inited Engineering Center 345 East 47th Street New York, N.Y.10017 i
,j
0 1
f' l
VIBRATION OF TUBE BUNDLES SUBJECTED TO AIR. WATER AND STEAM. WATER CROSS FLOW:
]
PRELIMINARY RESULTS ON FLUIDELASTIC INSTABILITY
..i.,,
F. Amise. B. Wlard, R. J. Gobert CENS/DEMT Department E twoe, Mecaniawei et Therm.aues Centre (tuden Nwcisa. ret Saciav. France G. Hetsrone Techn.on - israei inst.rute of Technology Haifa Israel F sun.e me, F r am atome l
Far.: La Oe' ease. France b
i
.(
ABST RFT lil This paper presents partial results concerning flow induced vibration h
in a
square pitch tu be bundle successively subjected to 6
3 air-water flow of equivalent denetty ranging from 4 to 100 kg a and then to 1
steam-water flow of de.altf ranging fram 40 to
-3 f
400 kg a The most prominent vibration excitation mechanism observed was s
'T
.7 e
fluidelastic instability. Ai r-wat e r and s team-wat er crit ical f low velocities i
t are adnaustely predicted using the classical Connors' proportionality factor of about 7. Turbulent excitation mechaniae of the tube bundle will be I
e.
reported in a subsequent paper.
1.
1NfKuDUCT10N e
.I Tubes of heat exchangers and other equipsent, may be damaged due to t
f, flow indured vibrations (FIV). These vibrations are caused primarily by r
g single-and two phase flows pe rpendicu la r to the axis of the tubes.
The 7
mechani Ms which induce the vibrations are:
fluidelastic instability and g
wort.: shedding, which causes laret amplitude vibrations; and turbulence, which causes small amplitude
- v. orations, t e r. o '
in fretting-wear and
'[
fatigue.
I lle ce nt reviews (1,2l reveal that
'm is i of the osst expe ritsental work has i
g y
be.e n done in single phase flow, and only limit a are availa51e on two
,h I phase air-water flows. Fluidelastic in,t ab t lit y has been obse rved in straight tube bundle {3,41, turbulence excitation and air-water flow regime effect4 y
l}f where me.asured in a straight tube hundle (S} and on a single tube (6}.
i di*
For practical appli :sti rs data basa cecerning FIV in two phase flow has sttil to be considerat,4v enlarged. Furthermore Flv mechanisms are still poorly undtrstood in rela:;on with flow regim s.
Finally j
1 b
availabiltty of air water mixtures to simulate properly steam-water flows in L
FIV experiments may still be ques t ionable. It is obvious ly necessary to test i
[
1 269 1?
I
!}
l r
I E
k her -
M tube length is:
Assuming the fluid added sass per unit m =8oe d8/4 (9) 1 a
3 w,ere one could infer from the data that the coefficient t.25852.0 Since the is small compared with the mass of the tube s,.
the added mass e, y
accur.cy of the added eass is not ve ry irportant - though it is in agreement 7
with the literature 18}.
g N
E' 4.3 Dasming Tube damping has been calculated from the half power at the frequency j
where maximum power occurs. Inese have been calculated only f rors turbulence excited vibrations. Since the me thod is not very accurata f.3 r low values i
of damping, the current data is scattered within a 150% range. One can make the following observations:
";o and the drag directions has similar values, without
-5 Damping in the lift is a clear dependence of damping on clear dependence on flow velocity. There steam-water mixtures jisplaying the 4
quality (figure 7) - whe re air-wate r and same trend 1.e. decrease of damping valueg from e
- 3 "
at x=0.01 to n
= 2=10' at x=1.
E n I c,(%)
e j
g c.
- FWil tint : Cartutti (91
'I
+ EV A 1 Steam-water data H3,i
--+-- Air-w ater tests 3 -
o EV A I kt-water data E
1
-e-Steam-water tests l
4
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2
'f
- c. Ff i p uit/ 24 0 02 s S> u.- G A
-Q- (== 9, f xi., 3 e
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+
, 4 = r. U
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i
{t 0 Ci
+
1, y
i X YHW
!,.)
i C + w- - - - -- _ _ - -
' ~~
0 02 04 06 08 1
f.
g 0.2 04 06 0.0 1
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Void fr action s.
-w-K y.
]
F i gii r.e 7: Damping in t-o-phase Figure 8: Two phase damping versus void friction yr flow c.
5 Results have been compared.av values from Carlucci & Brown (9], who i
=
g h..
dettne two phase dimping as foi m:
(17) g c p
- c, -( c, +c,* eg)
- N 3
F
]
i$
}
- O.2% Viscw4 43* ping c, where one can assume a structurai 4 3*f ind c,
l t.
i I
278 j
_j 1
1
=
3 M
:m
-..,,._,.u.
.m.
f e_