ML19320B089
| ML19320B089 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/01/1980 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8007090287 | |
| Download: ML19320B089 (14) | |
Text
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e iaQ4 ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK, ARKANSAS 72203 (501)3716 July 1, 1980 1-070-02 Director of Nuclear Reactor Regulation ATTN:
Mr. Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No. DPR-51 Outstanding Items Related to B&W Small Break Analysis (File: 1510.1)
Gentlemen:
Pursuant to the D.F. Ross letter of August 21, 1979, Arkansas Power &
Light Company herein submits a quantitative response to Item 2B.
At-tachment I quantifies the lowered loop 177 FA Babcock & Wi'cox plant 2
response to a 0.01 ft cold leg break without feedwater and subsequent repressurization and sticking open of the pressurizer power operated relief valve.
The analysis shows the present operator guidelines are adequate to mitigate the consequences of this postulated event.
Very truly yours, t w s e e. T David C. Trimble M.snager, Licensing DCT:DGM:nak Attachment M
f-8007090 MEMGEA MIDDLE SOUTH UT!UTIES SYSTEM
49 SFMLL ' BREAK WITH FAILED PORV
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3 1.
INTRODUCTION it has been established in reference 1, that very small cold leg breaks (<0.01) will repressurize to the PORV setpoint of 2465 psia if the auxiliary feedwater is delayed significantly.
Since there is a probaLility of the PORV sticking open af ter being actuated, concerns have been raised regarding the impact of this consequential failure.
This report presents the results of an analysis of a 0.01 ft2 cold leg break-with the subsequent failure of the PORV to close.
2.
SUMMARY
& CONCLUSIONS As has been demonstrated by the analyses presented in Section 6 of reference 1, small breaks in the primary system will not cause a repressurization to the PORV setpoint unless all feedwater is lost to the steam generators.
Under this 2
situation, there exists a class of very small breaks, (less than 0.01 ft )
wherein the system will repressurize to the PORV setpoint.
An analysis is pre-2 sented herein for a 0.01 ft break, without feedwater to the steam generator, which results in a repressurization to approximately the FORV setpoint. At 20 minutes, the FORV was actuated and was assumed to stick open.
As is ' demonstrated in Section 4, for the 177-FA lowered-loop plants, operator action by 20 minutes to manually actuate the two high pressure injecticn trains will keep the core covered. A qualitative analysis is also presented which demonstrates that reestablishment of auxiliary feedwater by 20 minutes, for
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both the 177-FA raised and lowered loop plants, will prevent core uncovery.
2 Therefore, a 0.01 ft break with no auxiliary.feedwater can be mitigated safely with B&W's present operator guidelines.
These estrator guidelines require establishing feedwater lo the steam generator as soon as possible, if the AFW is not available initially, and manual initiation of the HPI upon loss of the steam generator heat sink or saturated conditions in the primary system.
'3..
METHOD OF ANALYSIS Evaluations of very small breaks which result in repressurization phenomena are presented in reference 1.
These analyses demonstrate that if auxiliary ieedwater is del'ivered to the steam generators, the primary system would not
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repressurize to the PORV setpoint. However, the analyses in reference 1 also i
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d:monstrate that if feedwater is not delivered to the steam generator within 20 minutes, there is a class of very small breaks, less than 0.01 ft2, which will
.re.sult in system repressurization to the PORV setpoint.
Since the PORV might
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stick open after being actuated, concerns have been raised regarding the impact of this consequential failure.
An analysis of a 0.01 ft2 break in the cold leg pump discharge piping, without auxiliary feedwater to the SG, was performed wherein the PORV was actuated and assuned to stick open. As has been demonstrated in reference 1, larger breaks will result in automatic actuation of the HPI system and will not repressurize.
While smaller breaks will rep'ressurize to the PORV setpoint earlier, less in-ventory would be lost out the break.
Therefore, the 0.01 ft2 small break with the subsequent failure of the PORV is expected to be the worst case for t ran-sients of this type.
The analysis was performed using the B&W ECCS evaluation model for the 177-FA lowered-loop plants.2 The analysis was performed using the same model and assumptions listed in Section 6.2.1.3.5 of reference 1 with th9 only changes being those made to reflect the PORV sticking open.
Key assumptions of the analysis are listed below.
1.
The initial core power level is 102% of 2772 MWt.
2.
The core decay heat is based on 1.2. times the ANS standard.
3.
Operator action was taken at 20 minutes to manually actuate both HPI pumps.
4.
The PORV was modeled as a leak path on the top of the pressurizer.
The orifice area of.0073 ft2 was used, Fowever, a C f 0.72 was utilized in D
order to reflect the proper relief characteristics of the FORV with the Moody critical flow model.
5.
The PORV was opened at 20 minutes.
This is consistent with the operator guidelines for a LO'C'A with no feedwater to the steam generators.
- However, if the operator had not acted within this time frame, approximately a 2 minute delay in operator action would have resulted in the PORV being actuated automatically.
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4 4.
RESULTS
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Figures 1 through 7 show the system response during the transient and Table 1 presents a sequence of events for this accident.
The resultant system pressure 2
response of a 0.01 ft cold leg break with no AFW is shown in Figure 1.
This particular response is due to (1) the loss of the SG heat sink; (2) no automatic HPI actuation prior to the loss of the steam generator heat sink; and (3) the opening of the PORV and actuation of the HPI at 20 minutes.
As seen in Figure 1, the pressure initially decreases following the break opening.
During this de-pressurization period, the reactor trips, the pumps trip, the pressurizer empties, and the steam generator secondary inventory boils off.
With the loss of the SG
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heat sink, the primary system starts to repressurize before the ESFAS signal is reached.
Therefore, the RPI is not automatically actuated.
The system repres-surizes to 2350 psia by 20 minutes at which time the PORV was assumed to open.
This is only 115 psi below the PORV setpoint which would have been reached ap-proximately 2 minutes later.
However the operator is instructed to manually open the PORV if the system repressurizes and the SG heat sink is lost.
- Thus, the opening at 20 minutes is not totally arbitrary.
During the system repres-surization the pressurizer level increases (Figure 2) and when the PORV is opened the pressurizer rapidly fills with two phase mixture. At the time of the PORV opening, the two HPI pumps are manually ~ actuated, and due to the addition of the cold makeup water and the additional leak path area, the RCS depressurizes.
The inner vessel mixture height is shown on Figure 3.
As can be seen, operator action by 20 minutes to manually actuate the HPI prevents core uncovery and a ninimum two-phase mixture level of 4.5 feet above the top of the core is main-tained.
Long term cooling is established at 25 minutes as the injected HPI fluid exceeds the core koil-off.
Thus, the acceptance criteria of 10 CFR 50.46 are satisfied.
While the analysis performed herein addressed the effect of operator action to manually actuate the HPI by 20 minutes, the effect of operator action to manually restore the auxiliary feedwater within 20 minutes can be qualitatively asses.ced.
As has ocen shown in Section 6.2.1.3.5 of reference 1, actuation of the auxiliary feedwater system at 20 minutes for a 0.01 f t2 break results in a rapid system depressurization and the subsequent actuation of the HPI.
For the case analyzed herein, the depressurization effect of the auxiliary feedwater would be faster than that shown in reference 1 due to the effect of the loss of inventory through the PORV.
Thus, the HPI would be actuated earlier and long term cooling vould be established faster than that shown in reference 1.
Therefore, no core uncovery is expected if the operator cnly actuates the auxiliary feedvater sys-tem within 20 minutes and, contrary to the guidelines, does not manually actuate the liPI.
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Table 1.
Secuence of Events Event Time, s 1.
0.01 ft2 cold leg break occurs
- 0. 0 2.
Reactor trip, loss of feedwater, and RC pump trip 54.5 3.
Main feedwater coastdown ends 60.0 4.
SG secondary boils dry 270.0 5.
PORV opened 1200.0 6.
HPI is manually initiated 1200.0 7.
Long teri cooling established 1510.0 h
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REFERENCES I Letter J.H. Taylor (B&W) to S. A. Varga (NRC), " Evaluation of Transient Behavior and Srall Reactor Coolant System Breaks in the 177-Fuel Assembly Plant," May 7, 1979.
2 B.M. Dunn, et. al, "B&W's ECCS Evaluation Model," BAW-10104, Rev. 3. Babcock
& Wilcox, May 1975.
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Figure 1 2
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.01.FT2 COLO LEG BREAK W/ND AFW 2 HPI'S & STUCK PORY AT 20 MIN. - COLO LEG BREAK LEAK FLOW 00AllTY 1.200 1.000 r
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