ML20040A414
| ML20040A414 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 04/14/1976 |
| From: | Norian P Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210040 | |
| Download: ML20040A414 (5) | |
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APR 141976 Docket No. STN 50-561 Richard C. DeYoung, Assistant Director for LWRs, DPM THRU: Zoltan R. Rosztoczy, Chief, Analysis Branch, DSS
~Z 2 R FIRST ROUND QUESTIONS - B-SAR-205 Document Name: B-SAR-205 Docket No.:
STN 50-561 Licensing Stage: CP Milestone No.: 05-21 Responsible Branch & Project Manager:
LWR-3, T. Cox Technical Review Branch Involved: Analysis Branch Description of Review: Round One Questions Requested Completion Date: March 26, 1976 Review Status:
Incomplete The Analysis Branch has reviewed B-SAR-205, which is the standard
{ NSSS submitted by Babcock & Wilcox. The enclosed questions relate to
.he methods and assumptions used to calculate mass and energy releases from the break following a LOCA or main steam line break analysis. This N
information will be utilized to design the containment systems by the -- -
applicant for the BOP.
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Paul E. Norian Section Leader Analysis Branch Division of Systems Safety
Enclosure:
Questions:
cc:
S. Hanauer R. Heineman D. Ross Z. Rosztoczy T. Cox T. Greene D. Shum W. Jensen (L
8201210040 010403 PDR FOIA MADDEN 80-S15 PDR J
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f"" A First Round Questions
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B-SAR-205 220.0 Analysis Branch 222.1 With regard to the mass and energy release calculations for the (6.2.1) containment subcompartment analysis discussed on page 6.2-1, pro-vide the following information.
a). 'The version of the CRAFT code referenced for these calculations uses the Moody correlation to calculate the flow at the break.
We do not believe that this approach is sufficiently conserva-tive for the calculation of subcooled flow rates. We believe the modified Zaloudek correlation used in later versions of f'~
the CRAFT code is acceptable for this purpose. Using an ac-
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ceptable break flow model, provide mass and energy release rates in tabular form, from time zero to about 1.5 seconds at approximately.05 second internals, for the following pipe failures.
1.
double-ended hot leg; 2.
double-ended pump suction break, 3.
double-ended pump discharge break, 4.
double-ended pressurizer surge line, 5.
d.ouble-ended pressurizer spray line, 6.
double-ended flood tank line, 7.
double-ended steam line, 8.
double-ended feedwater lines.
l b.)
Using the above method, provide the results of a sensitivity study in which the noding in the piping adjacent to the break j
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is increased until a convergent solution of break flow is obtained. Justify the break noding of each of the above postulated piping failures.
c.)
The above analysis should be based on a reactor power of x
3800 MWt with an additional 2% increase in power level to account for instrument error.
222.2.
For the long term mass and energy release calculations discussed (6.2.1) in section 6.2.1.3, there are two sets of mass and energy release information. The first set in tables 6.2-6 through 6.2-18 appear to assume complete quenching of steam by the injected ECCS water.
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The second set in tables 6.2-19 through 6.2-22 assume no quenching
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of steam by ECCS water. Discuss which of these data sets are pro-vided as interface information to the BOP for use in sizing the containment safety systems. Discuss the purpose of the other data set.
If the above data without steam quenching is to be the interface 222.3.
(6.2.1) information to be used in the BOP design, provide the mass and energy release for a complete spectrum of pump suction breaks.
Also, include the mass and energy release data for double-ended break at the pump discharge. These analyses should be based on a power level of 3800 MWt with a 2% overpower allowance to account for instrument error.
222.4 For the double-ended pump suction break, provide the average core
'6.2.1) temperature at the end of blowdown.
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222.5.
In the long. term mass and energy release calculation of section (6.2.1) 6.2.1.3., the CRAFT code is used to calculate mass and energy releases in the reflooding period as well as the blowdown period.
6 3
You state that a containment volume of 3.4 x 10 ft is used to compute the containment pressure for the reflooding part of these calculations.
For the double-ended pump suction break, provide the containment a.
pressure calculated for the reflooding analysis as a function of time.
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b.
BOP containment designs may have higher containment pressures than the values utilized in your analysis.
During the reflooding period, higher containment pressures will produce less steam binding
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in the coolant loops which may increase the mass and energy release rate to the containment. Provide a sensitivity study showing mass and energy release to the containment as a function of assumed con-tainment pressure and discuss the effect on the containment pressure that would result from these releases.
222.6 During a LOCA when the primary fluid temperature decreases below (6.2.1) the secondary fluid temperature, reverse heat flow in the steam generators will produce boiling in the primary system which will provide additional steam to the containment. A reverse heat transfer multiplier of 0.1 was assumed in B-SAR-205.
Provide justification for this assumption by comparison with the appropriate l
heat transfer-correlations on the primary and secondary side of the tubes.
Include consideration of steam condensation in the upper SM regionofthesteamgenerator(.7econdaryjasaresultofthisprocess.
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-4 222.7 Compare the carryout fraction from the core calculated by the (6.2.1) CRAFT code during the reflooding period to the data from the Fl!CHT experiments and justify the values calculated by CRAFT.
222.8 The mass and energy release data in B-SAR-205 for subcompartment (6.2.1) analysis and' containment design basis analysis was calculated based 1
f on a core power of 3760 MWt and adjusted to a power of 3800 MWt.
Provide a detailed description of how this adjustment was made for both subcompartment and DBA analysis.
222.9.
The following questions concern the calculation of mass and energy i.
(15.1.14) release to the containment for a main steam line break described
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in section 15.1.14.
These calculations were performed using the TRAP code, a.)
Provide the assumptions made for steam separation in the w
secondary steam side of the generator. Discuss the conservatism of these assumptions for containment analysis.
b.)
Table 15.1.14-2 lists the initial energy content of one steam 6
generator at 29.93 x 10 BTU. Table 6.2-23 gives the initial fluid 6
6 energy content of two steam generators at 87.9 x 10 BTUs or 44 x 10 BTUs per steam generator.
Discuss this apparent inconsistency.
c.)
Auxiliary feedwater flow to the ruptured steam generator will provide an additional long-term source of steam to the containment.
Discuss how auxiliary feedwater is treated in your analysis.
d.)
Provide the input constants and assumptions used to calculate i
reverse heat flow in the intact steam generator.
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p UNITE 3 STATES o~
NUCLEAR REGULATORY COMMISSION
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j WASHINGTON, D. C. 20666 t
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APR 141976 Docket No. STN 50-561 r
Richard C. DeYoung, Assistant Director for LWRs, DPM THRU:
Zoltan R. Rosztoczy, Chief, Analysis Branch, DSS M R FIRST ROUND QUESTIONS - B-SAR-205 Document Name:
B-SAR-205 Docket No.: STN 50-561 Licensing Stage: CP Milestone No.:
05-21 Responsible Branch & Project Manager:
LWR-3, T. Cox Technical Review Branch Involved: Analysis Branch Description of Review: Round One Questions Requested Completion Date: March 26, 1976 Review Status:
Incomplete The Analysis Branch has reviewed B-SAR-205, which is the standard
( 'SSS submitted by Babcock & Wilcox. The enclosed questions relate to i
(
ne methods and assumptions used to calculate mass and energy releases from the break following a LOCA or main steam line break analysis.
This information will be utilized to design the containment systems by the cpplicant for the BOP.
l 1
N Paul E. Norian Section Leader Analysis Branch Division of Systems Safety
Enclosure:
Questions:
cc:
S. Hanauer R. Heineman D. Ross Z. Rosztocz T. Cox / y T. Greene l
D. Shum I
W. Jensen
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