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UNITED STATES NUCLEAR REGULATORY COMMISSION g
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JUN S 197 7 Docket No. STN 50-561 MEMORANDUM FOR:
- 0. D. Parr, Chief Light Water Reactors Branch No. 3, DPM FROM:
T. Cox, Project Manager, Light Water Reactors Branch No. 3, DPM
SUBJECT:
FORTHCOMING MEETING WITH BABC0CK & WILC0X (B&W)
TIME & DATE:
1:00 PM Monday, June 13, 1977 LOCATION:
Room P-422, Phillips Building Bethesda, Maryland PURPOSE:
To discuss B&W's current position and their status in preparing responses to the attached requests for information concerning steam generator flooding and -steam line break concurrent with generator flooding.
PARTICIPANTS:
NRC T. Cox, G. Mazetis, S. Newberry B&W J. Happell J. Hamilton E. Swanson V. Galan E. Oelkers AY omas i. Cof, Project Manager Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
7-Requests for Information
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l '. 1 ENCLOSURE 7
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RE0dESTS FOR ADDITIONAL INFORMATION
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TO BE DISCUSSED AT 06/13/77 MEETING WITH B&W b
March 17, 1977 I
1.
Provide probability estimates, with bases,' for the following e ents:
{e a).
OTSG flooding b).
Steam line break concurrect with OTEG flooding l
6 These estimates should include consideration of the various possible operating conditions and event sequences.
2.
Provide an analysis of each of the above events, including an identifica-tion of initial conditions, input assumptions and plots of key primary f
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and secondary parameters. Discuss the conservatism of the initial i
conditions (i.e., highest RC flow, burnup, etc.).
Include a table g
or depiction of the times of occurrence of each of the key protection actions and each significant point in the transient sequence (i.e.
transient initiation, GTSG maximum water level, feedwater isolation, etc.).
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l 3.
Estimate the amount of moisture or liquid carryover for the above events.
1 4
Discuss the potential dynamic effects that a high moisture or liquid r
carryover may incur. Relate especially to the systems or components c,
8 relied upon to mitigate the consequences of the above e ents.
3 j
s %t 5.
Discuss the potential for a turbine trip or a loss of offsite power 4
occurring as a result of a OTSG flooding event.
Indicate the effect 3
1 of such a perturbation on the flooding event analyzed in each of the above events.
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.6.
With regard to the steam line break concurrent with flooding of the e.
OTSC, provide an evaluation of reactor vessel thermal s:.resses in-cluding a fracture mechanics analysis. Specifically provide:
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$ 3 a).
Adeterminationoftemperatureprofilesinthereactorvesshl 11.
b).
The combined pressure and thermal stress profiles at various'elmes,
i in the transient.
s c).
A calculation of stress intensity factors based on stress profiles.
'j; d).
A comparison of the calcuiated stress intensities with the reference 1
fracture toughness curve. (EOL) l.
e).
A determination of critical and maximum arrest depths in the
,t reactor vessel wall.
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i, n s M_EETIrlG fl0TICES DISTRIBUT!0fl
,m Docket File J. Knight NRC PDR D. Ross Local PDR R. Tedesco TIC S. Pawlicki J
LWR #3 File I. Sihweil NRR Reading P. Check E. G. Case T. Novak
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D. Crutchfield Z. Rosztoczy R. Boyd T. Ippolity R. DeYoung V. Benaroyo D. Vassallo G. Lainas D. Skovholt F. Rosa R. Denise V. Moore F. Williams R. Vollmer 1
J. Stolz M. Ernst K. Kniel W. Gammill O. Parr G. Knighton S. Varga B. Youngblood R. Clark W. Regan l~~"-
T. Spets D. Bunch P. Collins J. Collins C. Heltemes W. Kreger R. Houston R. Ballard R. Heineman M. Spangler H. Denton J. Stepp ACRS (16)
L. Hulman L. Crocker H. Ornstein
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H. Berkow L. Dreher A
l Attorney, ELD,i Project Managap T. Coi'1 s
IE ( 3) i B. Faulkenberrf.sIEs'
.5 SD (7)
M. Rushbrook 4
Receptionist NRC
Participants:
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4 00CKE:P NO: SIN 50-561 VENDOR:
BABCOCK & WIILOX 03MPANY (B&W)
SudJELT:
SumARY OF MEETING TO DISCUSS STAFF REQUESTS FOR INE0RMATION ON STEAM GENERATOR FEEDWATER ADDITION EVWrS On June 13, 1977, representatives of B&W and our staff met in Bethesda to discuss B&W's planned responses to our requests for information. contains the requests discussed. B&W staff members explained their plans by describing the (1) analytical tools to oe used, (2) the safety criteria to be used in designing the analyses and evaluating the results, (3) analytical work now partly completed, and (4) their preliminary schedule for reporting to NRC. Enclosure 2 is an attendance list.
E,Swanson (B&W) described B&W's apphach to the concerns expressed in the h
x six staff questions. He said that secondary systems and equipment have previously been designed with the primary objective of preventing moisture carryover to the turbine. The new analytical cffort is designed to evaluate wnether there is any decrease in plant safety resulting from single. failures in systems or parts of systems affecting feedwater flow rates.
V. d3 alan briefly described the analog computer method, POWER TRAIN 4, that ic being used in the single failure analyses. He said that the code is essentially identical to that described in a topical report (BAW-10070) now being reviewed, T
the major difference teing the addition of a second steam generator. Analyses done so far using POWER TRAIN 4 have been for assumed single failures in the
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Integrated Control System (ICS) and in the feedwater system, 6%U Thus tar, numed failures have been system fpnctional failures, without regard to whar. specific equipment failures could have caused the system to malfunction. For example, feedwater flow is assumed to increase from 15 percent to 100 percent without examining the detailed causes of such failure.
B&W responses to each of the six requests for information were discussed:
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(1) S&W said that4the lad-cf component reliability data makes probability estimation unreasonable. They are analyzing feedwater addition events due to single failures, anticipating that such events willJo_tmresult 3 / '
unacceptable consequences. The B&W rationale is thatgceptable consequences can be shown for the assumed event, then there is no mecessity Ecr the cte'm 1im to evaluate the probability of occurence of that eventg bek concurrent with atc= generator-f4codincJy-00 ag mcd that an eff*
to mm the-probsbility of thece concur +ent--events Ms r^qi^t J
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(2) & (3) V. Galan described some preliminary results of analyses.
Of twelve separate analyses having different initial conditions, two were shown to result in 10 to 30 percent moisture carryover for periods of approximately five minutes. B&W expects to be able to provide additional results of these failure analyses N
by July 15, 1977.
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(4) B&W has contacted several arch ect-engineers and General Electric about the potential effects of Mile " potential dynamic effects" are readily defined (erosion, missiles), liquid carryover quelitative limits on allowable carryover appear to be difficult to establish.
(5)
V. Galan stated that the flooding transients analyzed to date have shown that the reactor will not trip, thus a turbine trip is not expected. He also pointed out that a loss of offsite power mitigates the effects of the flooding trarys,ient whenever the power loss occurs (during the transient) because W feedwater pumps lose power.
k (M R&W reiterated their intent to show that consideration of steam line break concur steam generator flooding was not warranted.
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evento arc un mlarad) i-L. sacugh - er- 'udc consideration of the-accincd cvent a ould provide an addition nt, in terms of a stree a ysis of appropriate components, for the inc lity of ondary side ruptures caused by liquid carryover from the ste rator.
. J. Taylor, B&W licensing manager, stated that the responses discussed in this meeting would be submitted on the BSAR-205 docket.
'Ihomas Cox, Project Manager Light Water Reactors Branch No. 3 Division of Project Management
Enclosures:
1.
Requests for Information Discussed at 06/13/77 Meeting 2.
Attendance List O
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