ML19316A702
| ML19316A702 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/24/1980 |
| From: | Salah S Office of Nuclear Reactor Regulation |
| To: | Phillips L Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8005230748 | |
| Download: ML19316A702 (5) | |
Text
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MM 2 41920 MEM0PA!40Uli FOR:
L.E. Phillips, Section Leader, Reactor Analysis Section, Analysis Branch, DSS FROM:
S. Salah, Reactor Analysis Sectinn, Analysis Branch, DSS
SUBJECT:
WESTINGHOUSE NES TRIP REPORT - f%RCH 6-7, 1980 The meeting was held in Pittsburgh on flarch 7,1980 with Westinghouse, NRC staff and NRC consultants (ANL) to discuss the review status of the topical reports WCAP-9500, 9401, 8966 and 9230.
In addition, Westinghouse and the staff discussed the status of the NRC steam line break cudit calcu-lations for Virgil C. Summer plant. This trip report only covers the meetings I have attended while at Westinghouse NES.
1.
Information Exchange on the review of WCAP-9230 (Report on the Consequences of a Postulated Mail Feedwater Line Rupture).
This meeting was primarily to clarify some of the questions the NRC consultants, Argonne National Laboratory (ANL), has so they could generate first round questions. These ANL questions will be combined with existing NRC questions and submitted to Westinghouse. lists the 25 questions ANL wanted to clarify.
Glen E. Lang and Hiroshi Fujishiro of Westinghouse spent approximately three hours to clarify each question. At the end of tae session questions nos. 5, 7,15,16, 23, 24 and 25 were not resolved. Westinghouse said they will answer these questions in their answer to the first round questions of WCAP-9230, Review of WCAP-8966 (Evaluation of Mispositioned ECCS Valves).
2.
Westinghouse said all the authors of this report (four of them) are not longer with Westinghouse, therefore, they would have difficulty answering the questions.
However, Westinghouse representatives said they would like to go through clari-fication of each question so they can attempt to answer the questions in the near future. Therefore Westinghouse and ANL members went through each question for clarification.
3.
Steam Line. Break (Virgil C. Summer Audit Calculations)
Doug Wood of Westinghouse and I examined the input and output data used by MARVEL code of Westinghouse and NRC IRT code. These input and output data were used for steam line break calculations. The results indicated IRT was not calculating the steam discharge from the intact steam generators properly.
Reverse flow from the (b
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L.E. Phillips steam generators and out of the break before the steam isolation valves closed was not calculated correctly. Therefore this portion of IRT has to be modified.
In addition Westinghouse provided additional input data to be used in the IRT 2
code. These input data consisted of coolant volumes in various reactor system components, Doppler reactivity coefficient with stuck rod configuration, average boron reactivity coefficient and reactor shutdown margin with a stuck rod.
In addition Westinghouse prov:ded output data from their calculation of the base case. We also discussed the : team generator model used in both codes (MARVEL and IRT).
We ended the meeting and decided to talk about our results after BNL modified the IRT code to account for the steam flow from the intact steam generators.
After the modification of IRT code we would incorporate additional input data which was provided by Westinghouse before discussing the results with Westing-house.
'\\s S. Salah Analysis Branch Division of Systems Safety
Enclosure:
As stated cc:
R. Denise Z. Rocztoczy T. Speis R. Martin L. Lois F. Odar Le Kintner S. Salah A
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O re t u!.i: r d ".i t - T e e f t t e. Lgture WCAP-92 M. Repo r t on tbc Can,'.ecy nees o' a Inf ormation Exchange Ques: tons *
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1.
- p. 2-3, para. 2: k'hy are the Reactor Cooling System (RCS) pu=ps Since there shut down before the high head saf ety injection pumps?
i is no loss of primary coolant, does the pressurizer beco=e solid after high head safety injection?
If so, asc there any adverse consequences?
Does the "no boiling" criterion mean no bulk d
2.
- p. 2-6, last para.:
boiling or no boiling anywhere in the RCS?
y 3.
- p. 3-5: L'hy is the cain feedwater pump trip on safety injection signal not listed in the sequence of events (e.g., p. 5-4)?
V 4.
Table 3-1 (cont.): 1Tay is the safety injection pump cutof f pressure for two trains higher than for one, and does this 1.sply over-l l
pressurization of the RCS?
M 5.
Fig. 5-14:
Explain the strange break flow profile.
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J 6.
- p. 5-11, line 3:
k' hat is A and where is it in Fig. 5-207 p
y 7.
Fig. 5-22: Why after the Loop 1 temperatures come together do they again separate from about 50 to 60 see?
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8.
Figs. 5-24 and 5-25: Why does Fig. 5-24 indicate a rapid rise in pressurizer level frem 200 to 1000 sec, when Fig. 5-25 shows pressurizer insurge is almost zero during the same period?
/ 9.
Figs. 5-23 and 5-31: What are the temperature and steam generator 1
pressure variations for the fourth loop?
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Fig. 5-31: Why does one of the steam generators neither blow down cc:pletely nor increase pressure after steam isolation?
l v' 11.
- p. 6-3:
The results s-ithout of f site power er.hibit decreased pea.
he: leg tdsperature when compared to the case with of f site power available." Why?
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- These ars "Round Zero* questions.
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- *ha t is the flow limiting device and can it s se t t Inr. be 12.
- p. 3-2:
changed?
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13.
- p. 5-2: What is meant by "best estimate' plant situation, how was it determined, and why?
What is the item no. 17 parameter and what are the I
[ 14.
Table 5.2:
results?
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15.
During steam generator blowdown, two phase flow oscillat' ion may What is the effect of this oscillation on the low-low occur.
level signal for reactor trip.
Did LOFTRAN provide for an effect of backpressure
)[ 16.
- p. 3-6, para.1:
on injected flow?
how does the operator know that a main feedwater v/ 17.
- p. 2-3, para. 2:
Could he make an error in responding line rupture has occurred?
that would worsen the accident?
What is the ef fect of ine main feedwater shut-g/' 18.
Figs. 5-6 and 5-14:
off on break f1o'w?
4 Do the " minimum trip rod worth" (p. 6-2) and "minimu= required
(/ 19.
shutdown margin" (p. 5-15) values assume a shutdown rod of maximum worth being stuck in the fully withdrawn position?
if one of the motor driven auxiliary feedwater 20.
In a 3-loop plant, pumps should fail, will the required minimum 380 gpm of feedwater still be delivered to the two intact steam generators in the interval before operator action is taken and the turbine-driven auxiliary pump is started?
1 if the feedwater pipe to steam generator no.
21.
In a 4-loop plant,
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breaks, and the auxiliary motor-driven feedwater pump for steam generator nos. 3 and 4 fails, does not much of the feedwater to
-3 stean r,cnerator nos. I and 2 pass ut the t r e s*(
an! trave a In-adequate feed ater flowrote situation?
y/'22.
What is the capacity of two motor-driven auxiliary feedwater pu=ps and the capacity of the turbine-driven pu=p?
What blowdown rate vs. quality data were generated to justify the
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23.
asse:ption that 20 percent quality break flow before reactor trip is consarvative?
/ 24.
Is the comparison shown in Fig. 4-5 valid for worst-case initial conditions?
If not, provide such a comparison.
p(25.
Did.the break flow become choked before reactor trip?
If it did, was the correct discharge flowrate e= ployed in the calculations?
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