ML19316A896
| ML19316A896 | |
| Person / Time | |
|---|---|
| Issue date: | 05/08/1980 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 8005280036 | |
| Download: ML19316A896 (5) | |
Text
_..
DISTRIBUTION:
Central FHes MAY 8' 1980 NRC PDR SSPB File NSIC C TIC /
TERA Mr. Thomas M. Anderson, Manager DEisenhut RTedesco Nuclear Safe,ty Department JRMiller Westin'ghouse Electric C6rporation JBerggren P. O. Box 355 EHylton Pittsburgh, Pennsylvania 15230 ACRS (1)
LLois
Dear Mr. Anderson:
SSalah TSpeis
SUBJECT:
REVIEW OF WCAP-9230 WHaass BPurple Additional infomation is needed to complete our review of Westin'ghouse Electric Corporation topical report WCAP-9230, " Report on the Consequences of a Postulated MainnFeedline Rupture". The requested Additional inform-ation is entlosed.
This additional information is needed by May 27, 1980 to meet our review schedule. If you cannot meet this date, please inform us within ten days after receipt of this letter of the date you plantto submit your response.
Sincerely.
ORIGINAL SIGNED BY1 JAMES R. MIM /
James R. Miller, Chief Standardization and Special Projects Branch Division of Licensing
Enclosure:
Request for Additional 3nformaYion cc: Mr. W. Spezialetti.
Westinghouse ElectMc Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 l
Mr. A. Ball Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230
/
II
{ &........
,p..fk.....S U U. 5. 2 8. 0 0).(p Orr,Ce >
suRNaue
- '9'en;.ka ? lRM.1.1.1.e c....
On,e.x.1189...
5/8!80....................-
-...-...1 5
l PaA : rORM 318 (9-76) NRCMO240 N U.S. GOVERNMENT PR,NT,NG Orf' ICE: len -289-369
- J
1 ADDITIONAL INFORMATION REQUEST FOR REACTOR SYSTEMS REVIEW OF TOPICAL REPORT WCAP-9230
" CONSEQUENCES OF A POSTULATED MAIN FEEDLINE RUPTURE 212.1 Figure 3-1: Why don't the motor driven auxiliary feedwater pumps feed the bottom steam generator?
212.2 Figure 2-1: Where exactly is the assumed rupture location and what are the consequences of it being closer to or farther from the steam generator?
212.3 The following reports are referenced..but they _have not been accepted by the NRC:
a.
Page 8-1: WCAP-7878 (possibly an obsolete version of LOFTRAN).
b.
Page 4-4: WCAP-7337 (possibly an obsolete version of FACTRAN).
c.
Page 4-5: WCAP-9813 (possibly an obsolete version of NOTRUMP).
Therefore, on what basis can the calculated results reported be considered acceptable?
212.4 Pages 5-2 and 6-1: For both the "best estimate" and " worst case" transients provide the reactor trip and safety injection trip delay times used in the analyses.
212.5 Page 5-16: Provide data concerning sensitivity of calculated results to the time during each transient that loss of offsite power occurs.
212.6 Page 2.5,' last paragraph: Why is service limit D of the ASME Code, Sect. III, invoked for specifying maximum allowable pressures?
NRC has not accepted use of service limit D; also, do not the safety valves easily allow use of a more stringent service limit?
212.7 Figures 5-9 through 5-16 and 5-29 through 5-32 (" worst case" results):
Provide the time dependent behavior of the following relevant parameters (where missing):
steam line and feedwater (main and auxiliary) flowrates, safety and relief valve flowrates (did they exceed the rated capacities?),
steam generator water levels (with and without ruptured feedwater lines),
heat fluxes not already given.(s tther average or hot channel), DNBR, core average temperature, hot channe, exit temperature, steam volume fractions in core, maxirium fuel and cladding temperatures, steam temperatures, feed-water temperatures, safety injection flowrate, break flowrate back pres-sure.
(An aiternative to providing these data would be explaining why the data are not relevant, or that a particular parameter remains constant at a certain value.)
212.8 Figures 5-9 through 5-16 and 5-29 through 5-32 (" worst case" results):
What is the rod reactivity worth versus rod position curve assumed in the l
calculations, and is the assumption conservative?
e
6 212.9 Page 4-1: LOFTRAN evidently does not simulate either the main or auxiliary feedwater system (WCAP-7907S); do LOFTRAN calculations assume a minimum amount of feedwater flow to-the steam generators? (NS-TMH-1802, S/26/78) 212.10 Where applicable, provide quantitative values for each of the worst case
)
assumptions (pp. 6-1 and 6-2).
Separate assumed errors from nominal values. Provide a quantitative value for the minimum required " shutdown margin" (p. 5-15, item d).
1 l
212.11 Page 6-1: Provide quantitative values for worst case assumptions of (1) valve discharge rates and response times including, for example, opening and closing times (delay times) for main feedwater, auxiliary j
feedwater, turbine,and main steam isolation valves, and steam generator and pressurizer relief and safety valves; and (2) reactor trip delay times, safety injection signal delay time, and delay time for delivery of any high concentration boron injection required to bring the plant to a safe shutdown condition.
212.12 Discuss in detail how radial and axial power distributions in the core were used to determine the amount of local boiling in the core following a feedline rupture. Were open or closed coolant channels used in the core thermal-hydraulic model? How were the coolant flow rates in the fuel channels detemined?
212.13 In WCAP-9230, it is stated that one of the Westinghouse criteria for acceptance of consequences of feedline break is that "the DNB ratio is such that there is a 95 percent probability thct the limiting fuel rod does not go through DNB, with a 95 percent confidence level."
Show that present results meet this criteria in the hot channel.
Provide time dependent heat flux and coolant flow rates for the hot channel.
212.14 According to Table 3-1, 4 loop PWR plants have smaller power operated relief valves than smaller 3 loop plants (179,000 vs. 210,000 lbs/hr.).
Explain why smaller power operated relief capacity is adequate for 4 loop plants.
212.15 Describe more in detail how an active or passive device is used to limit the auxiliary feedwater flow to two intact steam generators in three loop plants to a minimum of 380 gpm. Provide the same description for a four loop plant where the minimum auxiliary flow rate of 470 gpm is required for the two intact steam generators.
212.16 It is stated that the NOTRUM? code incorporates the following transient features:
a.
a momentum balance suitable for predicting time-dependent recirculation ratios and flows.
b.
a suitable slip flow model for the thermal-hydraulic ccnditions involved.
c.
natural and mechanical phase separation models, including counter-current flow modeling capabilities.
. Provide experimental verification to justify the use of these models.
212.17 Provide experimental verification of the feedline blowdown qualities calculated with the NOTRUMP code.
212.18 Provide experimental verification of the calculations performed to determine the water level and heat transfer models in the steam generator using the NOTRUMP code.
212.19 How do you calculate the fuel and clad temperature without knowing the fuel element axial heat generation rate? (Refer to Figure 4-2).
212.20 In analyzing the sensitivity of broken feedline area, discuss the effect o# trip delay, start up of auxiliary feedwater pumps,-time of baron injection, and time of loss of offsite power. What effects do these parameters play in the consequences of a feedwater line break.
How do you determine the worst case based on your results.
212.21 How was the uncertainty in the initial mass inventory of the steam generators determined? Discuss the justification in detail.
212.22 For cases with loss of offsite power; peak hot leg temperature from saturation is quoted. How does this temperature compare with the temperature of the coolant in the hot channel? How much boiling occurs in the hot channel and what is the effect of this boiling on the system pressure and coolant channel pressure drop.
Discuss the results in detail.
212.23 Discuss how the uncertainty in the steam generator level was determined. Justify the error band used in this report.
212.24 It is stated on page 5-4 that for the worst case feeding rupture,'
the peak temperature is 2.5 F below saturation and occurs 0
3238 seconds after the break. What temperature does this 2.5 F below saturation temperature correspond to? It is the peak hot leg temperature, or peak hot channel temperature?
212.25 Figure 5-14:
Explain the unusual break flow profile.
212.26 Figure 5-22: Why after the Loop 1 temperatures come together do they again separate from about 50 to 60 sec?
212.27 Page 5-4, item 3: During steam generator blowdown, two-phase flow oscillation may occur. What is the effect of this oscillation on the low-low level signal for reactor trip?
212.28 Page 3-6, paragraph 1: Did LOFTRAN provide for the effect of backpressure on injected flow?
212.29 Page 4-8, line 7: What, blowdown rate vs. quality data were generated to justify the assumption that 20 percent quality break flow before reactor trip is conservative?
\\
l l
. 212.30 Is the primary hot leg temperature comparison shown in Figure 4-5 valid for worst case initial conditions? If not, provide such a comparison.
212.31 Page 4-8, item 2.a: Did the break flow become choked before reactor trip?
If it did, was the correct discharge f iowrate employed in the calculations?
212.32 Figure 2-1:
It is not stated whether the feed line rupture is inside or outside containment. Would not the consequences be worse if the rupture were outside containment?
212.33 Figure 5-15: Why does DNBR always increase? Show DNBR after 32.sec.
At what axial location does the plotted DNBR occur, and is it the minimum DNBR7 Of particular interest is the close approach of T to T at 3000 sec.
avg sat 212.34 Page 5-9, item 14: Why is the transient less severe when auxiliary feed-water is delivered to only one steam generator? (There is less heat trans-fer area, which could be more important than reduced total purge volume.)
Also, can the flowrate to one steam generator be as high as to two?
212.37 Page 4-3, item 3: What DNB correlation was used in the LOFTRAN code?
(If the W-3 correlation was used, justification for its use at 2500 nsia should be provided.
(W-3 is not valid above 2300 psia.))
212.38 Page 2-3, paragraph 2: How does the operator know that a main feedwater line rupture has occurred? What assurance is there that he wouldn't make an error in responding that would worsen the accident?
212.38 Figures 5-6 and 5-14: What is the effect of the main feedwater shutoff on break flow?
212.39 Page 2-2:
In a 3-loop plant, if one of the motor driven auxiliary feed-water pumps should fail after feedline rupture, how much auxiliary feedwater is assumed to be delivered to the two intact steam generators in the interval before the turbine-driven auxiliary punp is started?
Show that the assumed value is conservative.
O e
_