ML18136A172
| ML18136A172 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Surry, North Anna  |
| Issue date: | 10/29/1979 |
| From: | Shum D Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| References | |
| TAC-12263, NUDOCS 7911130135 | |
| Download: ML18136A172 (7) | |
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OCT 2 9 1979 PSB ROG MEMORANDUM FOR:. G. Lainas, Chief Plant Systems Branch FROM:
THRU; Division of Operating Reactors D. Shum, System Analyst Plant Systems Branch Division of Operating Reactors E. Aden*sam, Section Leader, Section B Plant-Systems Branch Division of Operating Reactors *
SUBJECT:
VEPCO PART 21 NOTIFICATION ON CONTAINMENT PRESSURIZATION On September. 7, 1979, VEPCO filed a/Part 21 notification stating that the current containment pressure response analysis performed by Stone and Webster for North Anna 3 and 4 showed that the containment may be overpressurized resulting from a main steamline break accident inside containment if the auxiliary feedwater flow to the damaged steam generator were not isolated-within ten minutes following the accident.
In the current analyses, S&hl assumed that auxiliary feedwater flow rates were at the pump runout conditions whiGh were not considered in the original analyses at the PSAR stage.
The Surry Station Units 1 and 2~ North Anna Unit 1, and Beaver Valley Unit 1 are operating plants designed by S&W with containments similar to North Anna 3 and 4.
To determine whether a similar problem may exist at these operating plants, we mat with the 1 icensees, VEPCO and Duque£1ael> and S&~J on Septeiriber 12, 1979.
As a result of the discussions during the meeting, we bel"ieve that North Arma l containment pressure analyses and design appear adequate*, because auxiliary feedwater fl ow at pump runout.con di ti on was considered in the analyses.
Furthermore, orifices are provided in the pump discharge lines. to limit the auxiliary feedwater flow to the steam generators.
Hov,ever, additional analyses were requested for Beaver Valley Unit 1 and Surry l and 2.
The pl ant designs for Beaver Valley 1 and Surry l and 2 are similar not identical) in many ways (e.g., they have similar NSSSi, reactor power rating, containment design and size, containment post LOCA heat removal system and capability, steam generators, secondary side water inventory-, auxiliary feedwater system design configuration, etc.).
Therefore, S&~J stated that the containment pressure response, analyses
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09,.0ct~bet l, 1979, S'~W provid~d the prelifinary results of the analyses
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Two st~_amline break sizes at u1
"'~ent'1 PO\\'{~r levels '.(102%_, 30% and 0% oK reac!or power) were
,,..,:>"1 :~.--analyzed.
The calculated pressures.(for a 9urat1on of 1800 seconds "foliowing a pipe break),for all cases analy,zed were beiow the containQ ment design pressure of 45 psig.
vJe have also perfonned scoping calculatiqns (Enclosure 2) to estimate the magnitude of pote_ntial containment overpressur.E; and the impact of*
varyrng auxiliary feedwater flow rates supplied to the broken steam generator~
Our analyses were based-'on the mass and energy release data provided in the Beaver Valley l FSAR for. & double-ended main steamline break at 0% power, minimum.en,gineered safeguards, and three various auxi 1 i ar.v feedwa ter 1flow.'.rates ( approximate 720 gpm.
1440 gpm, *and 2160 gpm_~. The design flow rate is 720 gpm).
Our analyses show that:
1~
The initial blowdown due to steam generator fluid inventory (which causes an initial containment peak pressure of about
- 3/4 of the containment design pressure at about.60 seconds).
will n_ot cause the containment to be overpressurized.
- 2.
With the current conservative assumption that the auxiliary feed\\~ater will be discharged through the broken steam generator into the containment at a constant enthalpy of 1200 Btu/lb, the magnitude of po ten ti al containment overpressure wi 11 be cl.
function of auxi 1 i ary _feed\\'Jater fl ow rates.
- 3.
Even for the case with the highest auxiliary feedwater fl ow rate analyzed, the containment would hot be overpressurized jf the auxiliary feedwater flow.to the damaged steam generator can be isolated within tei1 minutes; followi_ng the pipe rupture. presents a comparison of some pl ant parameters for Beaver Valley 1 and North Anna 3 and 4. It should be noted that the reactor power ratings for the above three plants are similar.
However, North Anna 3 and 4 utilize B&l1 1s NSSS eq1,1ipped with two once... through steam generators, while Beaver Valley 1 utilizes a Westinghouse NSSS equipped with three steam generators.
Due to peculiarities in the NSSS design (e.g., reactor returns to 50% power following a main steamline rupture3 smaller once through steam generator with less mass inventory, etc.)
B&W plants in general require highe,r. auxiliary_.feedwater*;flovi.th'an'*other plants with similar power ratings. Therefore, B&~I has specified auxiliary feedwater pumps with significantly higher output capability.
Each auxiliary feedwater pump for North Anna 3 and 4 discharges through dedicated piping and a dedicated p~netfation directly into
~a~c~n steam gen~rai:or.
As a result, to1101 rng a marn sti=amtrne Ol'l'IC"~
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rupture,, the auxiliary feedwater flow to the broken steam generator will increase to the equivalent of the total flow of all the associated pumps operating it their runout conditions.
Where as for Beaver _Valley plant, all the pumps discharge into their associated steam g_enerator thaough a common header and a single penetration.
Following a main steamline rupture, due to the decrease in the system resistence, the tomal pump flow as well as the piping.friction will increase to some extent.
The piping friction will help significantly to prevent the pumps from operating at their ~unout conditions.
Based on our comparison of conta.inment pressure response as a function
_ of auxiliary fee.dwater flow rate and the above system comparison, we believe the auxiliary feedwater system piping and pump runout conditions.
are the most signifkant in determining whether or not containment design pressure.will be exceeded.
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Enclosures:
As stated cc w/enclosures:
D. Eisenhut R. Tedesco W. Gammill B. Grimes R. Denise t~. Butler
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D. Ziemann R. Reid E. Adensam J. Kerrigan D. Neighbors D. Wigginton D. Shum D. Shum, System Analyst Plant Systems Branch Division of Operating*Reactors
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ENCLOSURE 1 Assumptions:
.1.
Main steam non-return valve functions instantaneously to isolate the faulted steam generator break from reverse main steam flow.
- 2.
Slowdown rate of the faulted steam generator is calculated by S&W/W, feedwater flow control valve 9*oes to full open, zero pressure drop through the steam generator for 30%
and 102% cases.
- 3.
The two intact steam generators stay pressurized.
- 4.
Steam generators are isolated by controls sensing the break.
- 5.
Auxiliary feedwater flow is available within ten seconds, with run-out flow going preferentially to the faulted steam generator per system hydraulics.
- 6.
Heat transfer coefficient is constant, heat transfer vs. time per computer code.
(~1200 BTU/lb.)
- 7.
Containment backpressure is 20 psig for all cases.
- 8.
Single active failure is one CIB resulting in minimum safe-guards: i.e., failure of one-half of the containment sprays.
- 9.
No operator action is assumed.
- 10.
All auxiliary pumps operational.
Footnotes (1)
Initial Conditions (2)
Initial Conditions 10.4 PSIA, Service Water T = 86°F, Tc= 105° (SAT) 11.6 PSIA, Service Water T = 32°F, Tc - 105° (SAT)
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ENCLOSURE 3
- 1. Auxiliary Feedwater Pum2s Motor Ori ven Number Runout fl ow each Turbine Driven Number Runout Flow
- 2.
Arrangements No. Aux. feed penetration in SG Lineup Line losses in analysis
- 3.
Runout flow to damaged SG e
Beaver Valley North Anna 3/4 2
550 gpm l
900 gpm l
Pump discharger headered together.
Significant 1592 gpm 2
1100 gpm l
1500 gpm 3
Each pump is routed to each SG through a dedi-cated line and a dedicat-ed penetration.
Normal valve alignment is 1-MD pump/SG.
Turbine driven pump feeds both SG's.
(In analysis they assumec not-driven flow to good SG.)
Ignored
'\\, 2600 gpm