ML19323C347
| ML19323C347 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 05/08/1980 |
| From: | Peoples D COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8005150375 | |
| Download: ML19323C347 (4) | |
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one First Naf 4nal P!3:a. C5.ca03. li! no.s k 2 ' ',7 Address Reply Io Pest Othce' Sox 76'7
'g Chicago. Illinois 60690 May 8, 1980 l
l Mr. James G. Keppler, Director Directorate of Inspection and Enforcement - Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Zion Station Units 1 and 2 Response to IE Bulletin No. 80-04
" Analysis O f A PWR Main Steam Line Break (lith Continued Feedwater Addition" NRC Docket Nos. 50-295 and 50-304
Reference:
February 8, 1980 letter from J. G. Keppler to C.
Reed transmitting IE Bulletin No. 80-04
Dear Mr. Keppler:
Reference (a) transmitted IE Bulletin No. 80-04, " Analysis of a PWR Main Eteam Line Break with Continued Feedwater Addition."
This Bulletin required action to be taken by Commonwealth Edison Company with regard to its Zion Station.
Attachment A to this letter contains Commonwealth's response to this Bulletin for this station.
Please address any questions that you might have concerning this matter to this office.
Very truly yours, f
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4.O. L.
Peop1 g
Director of
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Nuclear Licensing OLP: WFN: rap attachment
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NRC O f fice of Inspection and En fo rcemen t -Division of Reactor Operations Inspection THIS DOCUMENT CONTAINS 3703A P00R QUAL.lTY PAGES
Attachment A Response To Items 1-3 of IE Bulletin No. 80-04 1.
Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact o,f other energy sources, such as continuation of feedwater or condensate flow.
In your review, consioer your ability'to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable af ter extended operation at runout flow.
Resoonse Commonwealth Edison has reviewed the containment pressurization analysis for main steam line break and the inclusion of additional water sources ooes not significantly contribute to the pressure response.
At Zion Station, both the feedwater regulator valves (including the bypass valves) and the feedwater isolation valves close from a safety injection signal which, for this case, comes from steam line conditins ano containment pressures.
This isolatica prevents any water from the condensate and main feeowater systems from entering the steam generator.
The only water injected will be from the auxiliary feedwater system.
All three auxiliary feedwater pumps have a combined runout capacity of approximately 2960 gpm, although the valves are currently throttled in the lines so that run outflow will not be achieved.
Following a containment pressurization from a steam line creak, the operatur checks for high radiation, etc. to determine that it is not a main coolant break.
Then he checks the four steam generators for wide range level and pressure to determine the broken loop.
The operator then isolates the auxilary feedwater to that loop.
Simulator training indicates that this will normally be accomplisned in less than one minute and should always be accomplished in much less than ten (10) minutes.
Commonwealth Edison has determined that the increase in pressre due to the entire throttled auxiliary feedwater flow being injected for 10 minutes and tne maximum feedwater flow for the 10 seconds prior to feedwater isolation will, at most, be 2 psi.
Based on the above, Commonwealth Edison has determined that the total pressure inclucing tne-auxiliary feedwater addition would be at most 40 psig which is suostantially below the containment design pressure of 47 psig.
2.
Review your analysis of the reactivity increase which results from a main-steam line break inside or outside containment.
This review shoulc consider the reactor cooloown rate and the potential far the racctur :: rsturn to power w.th the most i
+
. reactive control rod in the fully withdrawn position.
If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated the report of this review should include:
l a.
The boundary conditions for the analysis, e.g. the end of life shutdown margin, tne moderator temperature coefficient, power level and the nat effect of the associated steam generator water inventory on the reactor system cooling, etc.,
b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, c.
The effect of extended water supply to the affected steam generator on the core criticality and return to power, d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ration (MDNBR) values for the analyzeo transient.
Resoonse Commonwealth Edison has reviewed the assumptions made for main and auxiliary feedwater flow as they apply to steamline break transients.
The transient analysis was performed using the following assumptions:
1.
The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowaole shutdown margin.
2.
For the Condition IV breaks, i.e.,
double-ended rupture of a main steam pipe, full main feedwater is assumed from the beginning of the transient at a very conservative cold temperature.
3.
All auxiliary feedwater pumps are initially assumed to be operating, in adoition to tne main feecwater.
Tne flow is equivalent to tne rated flow of all pumps at the steam generator oesign pressure.
4 Feedwater is assumed to continue at'its initial flow rate until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumeo to continue et its initis: floy rate.
5.
Main feedwater flow is completely terminated following feedwater isolation.
m.
, Based on the manner in which the analysis is performed for Zion Station, the core transient results are very insensitive to auxiliary feedwater flow.
The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat transfer, which is the forcing function 4
for both th3 reactivity and thermal-hydraulic transients in the core.
The effect of auxiliary feedwater runout (or failure of runout protection wnere applicable) is minimal.
Greater feedwater flows during the large steamline breaks serve to reduce secondary pressures, accelerating the automatic safeguards actions, i.e.
steamline isolation, feedwater isolation and safety injection.
The assumptions described above are therefore appropriate and conservative for the short-term aspect of the steamline break transient.
The auxiliary feedwater flow becomes a dominant factor in determining the duration and magnitude of the steam flow transient during later stages in the transient.
However, the limiting portion of the transient occurs during the first minute, both due to higher steam flows inherently present early in the transient and due to the introduction of boron to the core via the safety injection system.
In conclusion, Commonwealth Edison and its vendor, Westinghouse have evaluated the effect of runout auxiliary feedwater flows in the core transient for steamline break, and based on this evaluation, have determined that the assumptions presently made are appropriate for use as a licensing basis.
The concerns outlined in the introduction to this bulletin, IE Bulletin 80-04, relative to, 1) limiting core conditions occurring during portions of the transient where auxiliary feedwater flow is a relevant contricutor to plant cooldown; and
- 2) incomplete isolation of main feedwater flow, are not representative of the Westinghouse NSSS designs and associated Balance of Plant requirements.
3.
If the potential containment overpressure exists or the reactor-return-to-power response worsens, provide a picposed corrective action and a schedule for completion of the corrective action.
If the unit is operating, provide a description of any interim action that will be taken until the proposed correctiv'e action is completed.
Response
Commonwealth Edison has determined that the potential for containment overpressure does not exist and the return-to-power response is very insensitive.to the addition of auxiliary feedwater.
The re fo re, no corrective action is required.
l 3703A
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