ML17275A863
| ML17275A863 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/04/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 8103110382 | |
| Download: ML17275A863 (10) | |
Text
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Docket No.,i 50-397 7
Washington Public Power Supply System ATTN:
Mr. R. L. Ferguson Managing Director 3000 George Washington Way P. 0.
Box 968
- Richland, Washington 99352 bcc:
TERA NRC/PDR L/PDR NSIC TIC ACRS (16)
Dist.
< Docket File'Eltiwala LB81 Rdg DEisenhut BJYoungblood DLynch MRushbrook RTedesco RVollmer
- TMurley, DRoss RHartfield, MPA VNoonan OELD OIE (3)
Subject:
Request for Additional Information Regarding the WNP-2 Design Basis Safety/Relief Valve Loads In the course of our review of your report entitled, "SRV Loads - Improved Definition and Application methodology for Mark II Containments," "we have identified a need for additional information.
Our request for this information is contained in the enclosure.
If your have any questions on this matter, please contact the Project Manager, M. D. Lynch, at 301/492-8413.
Sincerely, cc:
See next page
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Robert L. Tedesco, Assistant Director for Licensing Division of Licensing SX93129>S~
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R. L. Ferguson Managing Director Washington Public Power Supply System P. 0.
Box 968 3000 George Washington May
- Richland, Washington 99352 ccs:
Nicholas Reynolds, Esq.
Debevoise
& Liberman 1200 Seventeenth
- Street, N.
W.
Washington, D.
C.
20036 Richard g. guigley, Esq.
Washington Public Power Supply System P.
0.
Box 968
- Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Mr. 0.
K. Earle
,Licensing Engineer P.
0.
Box 968
- Richland, Washington 99352 Nr. Albert D. Toth Resident Inspector/MPPSS-2 NPS c/o U.
S. Nuclear Regulatory Commission P.
0.
Box 69
- Richland, Washington 99352
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021.0 CONTAINMENT SY i
MS BRANCH 022. 053 022.054 022.055 The Caorso test results discussed in your report on the safety/relief valve (SRV) loads, "SRV Loads-Improved Definition and Application Methodology for Mark II Containments,"
exhibi t some high frequency pressure spikes in the boundary pressure measurements during the initial phase of the air clearing transient.
Since these spikes were not observed in previous quencher test data (i.e., the German quencher
. tests), this phenomenon suggests that the specific type of quencher design may be important in determining the, characteristics of air clearing loads.
Accordingly, provide a detailed description of the WNP-2 quencher
- geometry, including a description of the hub design.
Compare the geometry of the WNP-2 quencher design with the device tested in Caorso.
Discuss any differences that may exist between them.'ndicate how these differences might influence air clearing loads.
The,Caorso test results discussed in the SRV report cited above, indicate'hat the size of the vacuum breaker on the SRV line is important in determining the reflood transient after valve closure and, consequently, the subsequent valve actuation loads.
Indicate whether the size, number and characteristics of the vacuum breakers installed on the SRV lines of the WNP-2 facility are similar to those of the Caorso plant.
If there are differences, discuss what effects you expect these differences may have on your facility.
State how these differences and their effects will be incorporated in your load definition for the WNP-2 facility.
The design values for the transient SRV loads in the WNP-2 facility are based on single valve, subsequent actuation data from Caorso in-plant tests.
These design values are then used in load cases involving multiple valve actuations based on the assumption that actuation of multiple valves occurs only for the first actuation of the SRY's.
State whether this assumption can be supported by -a transient analysis of the worst transient event expected in the WNP-2 facility. If this is not the,
- case, revise your load definition to consider the multiple valve effect on the design basis pool boundary loads.
Our concern is that the Caorso test results indicate that pressure loads from multiple-valve actuations are greater than those from single-valve actuations under similar first actuation conditions.
022.056.
022.057 In Figure 6.8 of the SRV report cited above, you indicate that the frequency spectrum of the desi gn pressure-time histories can hound the experimental'ressure-time traces at a statistical confidence level of 90 percent/
90 percent and also bounds the envelope of the single valve subsequent actuation pressure traces from the Caorso tests.
Provide similar comparisons for leaky valve (LY) first actuation data and for multiple valve actuation (MVA) data.
Our concern is that the distinct differences in the character-istic c in LV data (e.g., the frequency and amplitude) and the greater riumber of initial pressure spikes in MVA data.
Many of the Caorso subsequent actuation tests were conducted with one of the two vacuum breakers blocked.
You used the results from these particular tests to derive the design values of SRV pressure transients-022-14
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, 022.057 (cont'd) 022.058.
for the WNP-2 facility.
However, the maximum pool boundary pressure measured in the Caorso tests is from a subsequent actuation test with both vacuum breakers operating (i.e., Test 22A02), which we believe to be prototypical for ftark II plants.
The maximum measured value of the peak positive pressure is.8.7 psi and the mean value is 5.9 psi for subsequent actuation tests with only one vacuum breaker functioning.
They are 9.4 psi and 7.3 psi, respectively, when two vacuum breakers are functioning.
This represents a potential non-conservatism in the data hase used in the derivation of design values of SRV pressure transients for the WNP-2 facility. Accordingly, discuss this phenomenon and its effect on the data evaluation, i ncludina your derivation of the design, basis SRV loads.
In order to account for the di.fferences between the WNP-2 design conditions and the Caorso test conditions (e.g., the pool geometry, the number of SRY's and the initial pool temperature),
you used a pressure amplitude multiplier based on a correlation in the Design Forcinq Function Report (DFFR) to obtain the WNP-2 de'sign values for SRV loads.
This procedure involves the extrapolation of pressure amplitudes measured in the Caorso tests with respect to some parameter values (e.g.,
the SRV steam flow rate) to WNP-2 design conditions usi'ng the trends established in the DFFR.
Accordingly, provide justification for your position that the trends used in this extrapolation can be supported by available Caorso data.
022.059 022.060 The proposed vertical pressure distribution in the SRV report cited above, is constant between the bottom of the suppression pool and the quencher elevation and then decreases linearly to zero at the pool surface.
You state that you derived thi s particular spatial variation hy reviewino'he maximum pressures measured at various elevations in the Caorso tests,
'owever, as shown in Figure 3.8a of the cited report, this proposed pressure distribution cannot hound the maximum measured values of pressure for all Caorso tests.
Furthermore, the use of the maximum measured pressure values in the comparison cannot reveal the effect of bubble. vertical motion on the measured pressure distribution.
Our concerns are that the bubble vertical motion will result in a more severe pressure distribution in the later part of the SRV transient and your model may not yield the correct pressure distribution.
Specifically, the cross-correlation coefficient of pressure traces measured at different elevations is less about 0.9 which indicates that there may he some effect from bubble motion on the pressure distribution.
Accordingly, modi-fy your proposed vertical pressure distribution, as required, to assure conservatism in the design load specifications for SRY transients-in the WNP-2 facility.
While you discuss both the proposed circumferential and vertical pressure distributions in the report cited above, you do not present a detailed discussion of the radial pressure distribution.
Accordingly, indicate how you calculate the radial pressure distribution.
State whether you assume the pressure distribution in the radial direction from the reactor
'edestal to the containment walls is constant.
If not,*describe the method you used and demonstrate that your approach is supported hy the measured pressure distributions in the Caorso tests.
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.922.061 You use the met~ odology in the DFFR to establish he 'circumferential
~ pressure distribution for the Wh!P-2 facility.
State whether'you use the line-of-sight and square root of the sum of the squares (SRSS) assumptions of the OFFR in your calculations.
If so, provide justification for usinq these assumptions in the.WHP-2 facility.
Your justification should be based on the Caorso test results.
Indicate to what extent these assumptions affect the WNP-2 load cases.
Provide representative figures showing the pressure. distributions on the basemat, the pedestal wall and the containment wall for the various SRV discharoe cases considered in WHP-2 plant design assessment.
022.062'ndicate what two valves are selected for the two-valve discharge case.
State whether the two quenchers selected are in the inner or the outer circle.
Provide justification for the two valves selected and for their location.
022.063 In the your analyses of SRV transients in the WNP-2 facility,.you assume that the pool water -is incompressible.
Your only justification for this assumption is that the cross-correlation coefficients between'ressure time histories measured at different locations are hioh.
Our concern is that this is insufficient justification since the relationship hetween the cross-correlation coefficient and the time phase shift has not been established; this relationship will influence the e'ffect of compressibility on pressure measurements.
You use the incompressible flow assumption in your analyses addressing the fluid-structure interaction (FSI) effect and in the WNP-2 structural analyses.
Even though the incompressible flow assumption can be justi'fied for the Caorso plant, it is still questionable whether it holds true for the WHP-2 facility.
Specifically, our concern is that the fluid-structure coupling effect may be more significant in the WHP-2 plant which has a steel containment than in the Caorso facility which has a concrete containment.
- Further, the velocity of sound in water is greatly reduced by the presenc'e of air and steam bubbles in the water; the conditions in the WHP-2 facility may differ'o the extent that the amount of air and steam bubbles in the pool water will be significantly different for the two fac'ilities.
Accordingly, since your assumption reaarding the incompressihlity of water in your analyses of SRV transienK in the WNP-2 facility is important but not adequately supported, provide additional justification on this matter.
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