Text

Forwards Response to 811113 Request for Addl Info Re Efficiency of Recirculation Pump Trip,Probabilistic Risk Assessment
	ML20039B741
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/17/1981
From:	Bordine T; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield D; Office of Nuclear Reactor Regulation
References
	NUDOCS 8112230505
	Download: ML20039B741 (18)
	v • d • e

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C0mpany s-DEC 2 219817 -

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D December 17, 1981 gg Dennis M Crutchfield, Chief Operating Reactors Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE'DPR BIG ROCK POINT PLANT - RESPONSE FOR ADDITIONAL INFORMATION REGARDING EFFICACY OF RECIRCULATION PUMP TRIP - PROBABILISTIC RISK ASSESSMENT By letter dated November 13,.I981, NRC requested additional information concerning the Probabilistic Risk Assessment results relating to the efficacy of a recirculation pump trip at Big Rock Point.

Previous information requests were made by the staff by letter dated August 5, 1981 and by subsequent telephone conversation.

Consumers Power Company letter dated September 10, 1981 provided our response to your August 5, 1981 request. The enclosure to this letter provides our specific response to the four additional questions raised in the November 13, 1981 letter.

QoDI

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Thomas C Bordine Staff Licensing Engir.eer CC Administrator, Region III, USNRC NRC Resident Inspector-Big Rock Point Attachment 8112230505 811217 PDR ADOCK 05000155 P

pop oc1281-0020bl42 a

f ADDITIONAL INFORMATION REGARDING EFFICACY OF RECIRCULATION PUMP TRIP PROBABILISTIC RISK ASSESSMENT nu1281-0020c142

f 1

Question #1:

In regard to the inadvertent control rod withdrawal as an ATWS initiator, provide simplified calculations to identify the following specific information:

(a) Power response of the system (b) Pressure response of the system (c) Time available for the operator to initiate the Liquid Poison System (LPS) in order to prevent actuation of the Reactor Depressurization System.

This information should be furnished for cases with and without recirculation pump trip.

Information should be presented in the format of page 82 of of the CPC submittal of February 26, 1981 and should include the effect of uncertainties in initial volumes as well as any time delays for the operator to be made aware that a transient is in progress.

Response

Ihe attached table presents the results of a hand calculation of plant and operator response to an inadvertent rod withdrawal from 100% power. The analysis assumes that reactor power ultimately reaches 336 mwt (140% Power) as presented in our submittal of May 4, 1981 on SEP Topic IV-2.

This represents a worst case analysis rather than an analysis which would result from a median or average worth control rod.

The analysis further assumes that the reactor does not SCRAM and for simplicity that the turbine bypass system is capable of accepting the additional steam flow resulting from this rod withdrawal without significant purtubation in reactor pressure. Cons is tant with our submittals of February 26, 1981 and September 10, 1981 the following assumptions are made.

A.

Initial Conditions 1.

Hot well inventory 25000 2500 lbm 2.

Drum inventory 24000 900 lbm 3.

Net primary system depletion required to actuate RDS 35000 lbm 4.

Initial Power 336 mwt B.

Low Level Transients 1.

Loss of feedwater results in a near immediate drop to 55% of the previous power level due to a loss of core inlet subcooling.

2.

Tripping of recirc pumps results in a power reduction to 50% of the previous power level.

h 2

3.

On depletion of the drum, 100 seconds remain until RDS actuation if RPT has occured (ATWS from 100% power was assumed to require 140 seconds to RDS af ter drum depletion in our 2/26/81 analysis.

From 140% power,140 sec (240 mwt)= 100 see) 330 mwt C.

High Pressure transients with feedwater from hot well 1.

Steady State power after SRV lifting levels out at 907. initial power.

2.

Feedwater trip occurs af ter depletion of 25,000 lbm from the hot well.

3.

Loss of feedwater results in a power reduction to 55% of the previous power level.

4 Loss of recire pumps results in a reduction to 607. of the previous power level.

5.

On depletion of the drum 100 seconds remain until RDS actuation if RPT has occured.

D.

High Pressure Transients without Feedwater 1.

Similar to high pressure transients with feedwater except feed pump trip and a reduction to 557. power occur early in the transient.

E.

Infinite Feedwater Transients l.

Full power is being relieved to the main condenser. Hours exist in which the operator can respond to initiate poison injection.

F.

Poison Injection Time 1.

With Recirc Pumps Running (41 occonds) 30 second purge time to establish a siphon plusan 11 second transit time through the pump, plenum and core.

2.

With Recirc Pumps Tripped (62 seconds) 30 second purge time to establish a siphon plus a 32 second transit time through the plenum and core (45/1.4).

The attached table presents the uncertainty in operator response time to inject poison resulting from calibration accuracy of steam drum and hot well level control instrumentation. The operator response time during low level transients should be shortened by 25 seconds to account for the short period of time required to deplete the drum to the low level SCRAM set point (35/1.4).

Plant and Operator Response Time f or Cont inuoug Rod Withdraw.nl (Worst Case Rod ) and Simult aneous Failure to SCRAM Inittat Power Power Time to Operator Power prior to after Drum Nos Power After Draus Hass Time To tst a time to inject Trinstent (Btu /hr x 10 6)

WPT RPT at RPT (lldn) insa of tV O Ines of FM RDS(sec )

1.PS Poison e

inw levet N RPT 1150 630 630 24,000 (1900) 108(13) 41 67(13)

RPT @ 60 seconde 1150 630 315 4550(1900) 630 24,000 (1900) 188(16) 62 126(16)

RPT 0 35 seconds 1150 630 315 12660(1,900) 630 24,000 (1900) 213(16) 62 152(16)

High Pressure wit h Feedwater N RPT 1150 1035 570 24,000 (1900) 181(13) 41 140(13)[t5]

[t5)

RPT O 60 seconds 1150 1035 340 21,600(1900) 570

?4,000 (1900) 294(16) 62 232(16)(18)

[18) f RPT G 8 seconde 1150 1035 620 24,000(1900) 340 24,000 (1900) 330(16) 62 268(16)[18)

[18]

I liigh Pressure M EM kPT @ 60 seconds 1150 570 340 7900(1900) 570 24,000 (1900) 209(16) 62 147(16) kPT G 0 seconds 1150 1150 340 24,000(1900) 340 24,000 (1900) 250(16) 62 188(16)

( ) - effect due to uncertainty in steam drum initial water level

( ) - ef fect duo to uncertainty in hot well initial water level to

F L

Question #2:

As a limiting calculation of the effectiveness of auto RPT in enhancing the operability of a non-qualified LPS, provide a calculation in the format of Table 1 of Attachment 2 of the CPC submittal of February 26, 1981 which includes auto RCP, an environmentally qualified LPS and an environmentally qualified Emergency Condenser System.

Response

Table 1 of Attachment 2 of our February 26, 1981 submittal has been regenerated to include three states of plant modification; the core damage probability without plant modification, with environmentally qualified poison and emergency condenser systems, and with an auto recirculation pump trip as well as the qualified poison and emergency condenser systems.

The automatic recire pump trip is modified as described in modification 2 ot Attachment 2 of our February 26, 1981 submittal. The core damage probabilities presented reflect operator action response times corrected for delays which may occur in making the operator aware that a transient is in progress. An automatic recirculation pump trip appears to reduce the core damage probability due to ATWS less than 207. as was concluded in our submittal of February 26, 1981.

TAliLE 1 SUt9-1ARY OF EFFECf OF ATW3 MODIFICATIONS 11% Pit Spurious loss or loss of Ioad loss of bit ri ent ions Fail closed TBPV FW 1FWP fieject Cond.

IICP MISC TiffAI.

-I

-6

-0

-I O.

Nane 3.2 x 10-3.3 x 10 2.8 x 10 2.9 x lo 8.0 x 10

6. 3 x 10 h.6 x 10 1.7 x lo 2 9 x 10 '

-6

-7

-6

-0

-6

-7

-I

-7

-5 1.

ENV. QUAL 2.4 x 10' le.8 x 10

2. 3 x 10 3.1 x 10 5 2 x 10 6.6 x 10 T.3 x 10 1.7 x 10 1.2 x 10 Ll"J & ECS l

-I

-6

-0

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ENV QtlAl.

1.6 x 10-2.h x 10 1.5 x 10

1. 3 x 10 5 2 x 10-2.8 x IO T.3 x lo 1 7 x 10 3,o, 39 1.l"J & ECS PLUG LOD Alfr0 !<l"r w

9 l

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l' age 1 of 4 Table 2 TA hlf: ?

FNVilutNME:NTAI, ytIAlJ FICATItill OF IIS & M'.I

~5 (Total-1.<' x IO /yr)

Sequence Transient Core Lumage Core Damage Frequency Frequegey Tranalent Seluen w Quantirication (Yr-1)

(Yr- )

-6 4

I t'h /l'tt Failure T AY L L

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T AY Lj:

Successful Turbine bypass

.66 Failure or reedwater given successful Turbine bypaas (Y )

1.0 g

k nual PHT within lat 60 sec.

9 Failure to begin poisun injection within 177 sec.

(212-35)

.37

-7 T AY.0L

(.18) ( 3.5 x 10 ) (.86) (1.0) (.1) (.92) 5.0 x 10 y

Failure of Manual hl'r within 60 a.ec.

.1 Failure of manual poison within 69 sec.

(104-35)

.92 T Ab I'

(.1 )( 3.5 x 10 )(.14)(.9)(:92)(10 )

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Page 3 of k Table 2 Uequence Trunstent Core Lamage Core I>amage Frequyncy Frequency (Yr* )

(Yr-1)

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Page 1 of 2 Table 3 TABLE 3-ENVIRONMENTAI.3tlALIFICAT10N f,PD & EC".

& MODIFIED AUTO PJ"P (0.97 x 10"

/ yr)

Cequence Transient Core Damm re Core Damage e

Frequency Frequency Trans ient bluence Quanti fication (Yr-1)

(Yr-1) 1.6 x 10" II'H/PH Failure T AY L

(.18)(3.5 x 10~ )(.86)(1.0)(1.0)(.29) 1.6 x 10' y

Failure to manually inject IIG at 202 see.

.29

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Provide specific data docu=enting a comparison of the core response (power) predicted by the RE2RAN codel and the BRP core si=ulator model, and the co=parison of the BRP core simulator model with the observed para =eters in the BRP core.

Eesponse:

Results of a co=pariaen of the core power response as predicted by the ETMN codel and the BRP core si=ulator codel are su==arized in the followir4 table. This comparison indicates very good agreement between the BRP core si=ulator codal and the REIRAN point kinetic:: model. Core thermal-hydraulic conditions (i.e. pressure, subcooling, flow) as predicted by PSIRAN at specific points in time were input to the BRP core si=ulator codel. End-ef-cycle 17 (all rods out) core conditions were assumed in the simulator codel as in the REIRAN :odel. Initial conditions for the core si=ulator 'codel were selected to duplicate the initial conditions assumed in REIRAN. Xenon was assumed to remain consta:1 throu6hout the transient.

The time points selected for co=parison were 60 seconds after initiation of a loss of feedwater ATWS without RFf, and 100 seconds after initiation of a loss of feedwater ATUS with RPT. At these times thermal conditions were charging, but only ver/- slowly. Because the BRP core simulator is a steady-state model, near steady-state conditions are required in order for the co=parison with FITRAN to be valid.

TABLE COMPARISON CF CORE PJ,lER RESPO!:3E AS PREDICTED BY RSTRAN AND THE BRP CORE SICL1 TOR POWER CASE SUBCOOLI G FLOW RETRAN CORE SI!6JLATOR Initial Condition 24.5Stu/lbu 100%

100%

100')

Loss of Feedwater ATWS 3.2 98.1 61.3 61.5 w/o RPT (t-60 sec)

Loss of Feedwater ATWS 1.1 42 5 28.1 29 5 w/RPI (t=lCO sec)

(

13 Comparisons between the BRP core simulator model and observed core parameters may be found in the " Big Rock Point Physics Methodology Report", dated August 25, 1978. This report was submitted on October 30, 1978 via a letter from DABixel (cPco) to DLZieman (Ima).

Question $4:

Provide additional clarification of your stat::nnt regarding the differences in initial conditions and safety valve setpoints for the analysis presented in NEDE-21065 and tv current analysis.

Response

It seems that I@0-21065 may be somewhat misleading as to assumptions made. Tables 3-1 and 3-2 list importaat init'ial conditions and equip-ment performance characteristics used in the analysis of the plant response to an AT4S event, including a reactor operating pressure of 1335 psig and a safety valve setpoint range of 1535 to 1585 psig. These values correspond to the canner in which' Big Rock Point is currently operated. Ecwever, these values were not explicitly assumed in the analysis of the NSSS power response reported in Section 6.2 of IE0-21065 The NSSS response reported in Section 6.2 is based upon an analysis of AT4S conducted prior to plant startup. This analysis is contained in Reference 1 to Iri:DO-21065 which is: " Transient Analysis Consumers Power Co=pany Big Rock Point Plant", General Electric Company, October 1962 (APED-4093). Pefering to this report it may be found that the 1962 analysis of AT4S (then called the safety valve sizing event) assumed an initial reactor operating pressure of 1500 psia and a safety valve setpoint which was 200 psi above the assumed operating pressure. For IEDO-21065 the NSSS response in an ArdS event was not reanalyzed other than to say that because the safety valve set pressure is still 200 psi above the NSSS operatir'g pressure, the rise in pressure, as well'as the peak heat flux

(% of rated), for the plant as currently operated should be no greater than that calculated in 1962. This conclusion is believed to be valid due to the fact that the void coefficient used in the 1962 analysis was extrencly censervative (i.e. very ne;;ative) as ccapared to the recent ccrac.

Tnis difference is believed to be dt.e to a number of factors including

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<A 1h the use of steel" channels vs the present zircaloy channels, a 12 x 12 '

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