ML19317F623

Responds to NRC 770607 Request for Addl Info Re Postulated Main Steam Line Break Accidents for Facility.Info Encl

ML19317F623
Person / Time
Site:	Davis Besse
Issue date:	08/09/1977
From:	Ror L TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
343, NUDOCS 8001220908
Download: ML19317F623 (17)

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i NRC PEM 195 V.S. NUCLEAn nEGULATonY CCMW'*SICN OCCKET NUM"E R.

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NSC DISTRIBUTION roR PART 50 DOCKET MATERIAL

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OATE CP OCCUMENT Toledo Edison 8/9/77 j

Mr. John F. Stolz Toledo, OH e,73,,f9/77

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!,b. Ci DISTRIBUTION OF MATERIAL CONCER'iING ONSITE EMERGENCY POWER SYSTEMS (1-P)

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Davis-hesse Nuclear Power Station Unit 1.

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%mm EDISON Docket No. 50-346 LOWELL E. RCE v.c. a m e.a.

Seria1 No. 343

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August 9,1977 fi AUGS 1977,d Director of Nuclear Reactor Regulation A W8e Attention:

Mr. John F. Stolz, Chief 4

Light Water Reactors Branch No. 1 Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Stolz:

Enclosed with this letter are the responses to your June 7, 1977 request for additional information regarding analysis of postulated main steam line break accidents coscerning Davis-Besse Nuclear Power Station Unit 1.

This submittal is in accordance with the schedule for responses stated in our 7/1/77 letter (Serial No. 312).

Yours very truly,

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Enclosures bj e/5 THE TCLECO ECISCN COMPANY EDISCN PLAZA 300 MACISON AVENUE TOLECO. CHIO 43652

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.' 6.44a.

Provide single active failure analyses which specifically identify those safety grade system and comporents relied upon to limit the mass and energy release and containment pressure / temperature response.

The single failure analysis should include, but not necessarily be limited to: main steam and connected systems isolation; feedwater, auxiliary feedwater, and connected systems isolation; feedwater, condensate, and auxiliary feedwater pump trip, and auxiliary feedwater run-out control system; the loss of or availability of offsite power; diesel failure when loss of offsite power is evaluated; and partial loss of containment cooling systems.

RESPONSE

A main steam line break that occurs inside containment must be a break of seismically qualified piping. In accordance with NUREG-0138 and NUREG-0153 and assuming a single active failure of a safety-grade component, credit may be taken for non-safety grade components as back up based on the reliability of these backup components. The safety grade systems and components relied upon to limit the mass and energy release from a steam line break (SLB) are the reactor protection system (RPS), safety features actuation system (SFAS),

steam feedwater rupture control system (SFRCS), control rod drive system, main steam isolation valves (MSIV), main feedwater isolation valves (FWIV), aux-111ary feedwater system, high pressure injection system, secondary side safety valves, and turbine trip via SFRCS actuation.

The single failure considered in Section 15.4.4 was the failure of the main feedwater isolation valve on the affected steam generator. This failure allowed the additional inventory between the FWIV and the feedwater control valve to be added to the inventory already available to be released. There is no single failure of the RPS, SFAS, secondary safety valves, turbine trip via ESFAS, or control rod drive system that could cause these systems to fail to meet their required function.

Similarly, the high pressure injection and auxiliary feedwater systems have been evaluated and shown that no single failure will defeat their intended functier..

(Ref: Response to Question 15.4.8).

The analysis presented in Section 15.4.4 assumes offsite power is available. Loss of offsite power, even with failure of one diesel, produces less severe core conditions and a small effect on mini =um suberitical margin (Ref. Response to Question 15.4.1). If one diesel fails instead of the FWIV, the mass and energy release to the containment is reduced as the aforementioned safety systems will still operate.

Therefore, the single failure analysis is reduced to the consideration of failure of a MSIV or a FWIV, with backup isolation provided by turbine stop valves and feedwater control valves, respectively.

If the mass of steam between the MSIV and the turbine stop valves is compared, in terms of mass of water, to the mass of water contained between the FWIV and the feedwater j

control valve, it is cle-that the worst single failure is the failure of

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the FWIV. This single failure releases the most mass and energy to the containment and produces the most severe core conditions.

Eingle failure analysis of the containment vessel heat removal syster-is discussed in FSAR Table 6-7.

b.

Discuss and justify the assumptions made regarding the time at which active conts h ant heat removal systems beccme effective.

1

D-B

RESPONSE

The accident analysis assumes that the containment heat removal systems (containment air coolers only) became effective 45 seconds following rupture.

Since two containment air coolers are normally operating, the greatest potential for a delay in their effectiveness will exist when offsite power is lost coincident with a steam line rupture. The total time for the air coolers to be effective is 43 seconds.

The components of this time are as follows:

Sequence of Events Elapsed Time Steam line break / loss of offsite power / diesel starts O sec.

SEAS setpoint reached / diesel up to speed 10 sec.

SFAS response time 15 sec.

Diasel sequence step 5/ containment air cooler starts 35 sec.

Containment air cooler accelerates to speed 43 sec.

This is within the 45 seconds assumed in the analysis, c.

Discuss and justify the heat transfer correlation (s)

(e.g.,

Tagami, Uchida) used to calculate the heat transfer from the containment atmosphere to the passive heat sinks, and provide a plot of the heat transfer coefficient versus time for the most severe steam line break accident analyzed.

RESPONSE

The Uchida heat transfer coefficient is employed to calculate the heat transfer from the containment atmosphkre to the passive heat sinks.

This is a coadensing steam value dependent on the ratio of water vapor to air.

Values of heat transfer coefficient are plotted against time on Figure 1 for the double-ended main steam line break.

d.

Specify and justify the temperature used in the calculation of condensing heat transfer to the passive heat sinks; 1.e.,

specify whether the saturation temperature corresponding to the partial pressure of the vapor, or the atmosphere temperature which may be euperheated was used.

RESPONSE

The saturation temperature corresponding to the partial pressure of the vapor was employed to calculate the condensing heat transfer to the passive heat sinks. See section 15.4.4.2.3 of the FSAR for a more detailed discussion.

e.

Discuss and justify the analytical model including the thermo-dynamic equations used to account for the removal of the condensed mass from the containment atmosphere due to condensing heat trans-fer to the passive heat sinks.

2

.D-B

RESPONSE

When the heat sink surface temperature is below the r:aturation temperature of steam at the partial pressure of vapor in,the conta bmant, water is condensed on the heat sink and transferred directly to the sump at the end of the calculation interval at the wall temperature. The equations are:

cond "

h,,e - hyg

( = A h (T - Y )

g where M

= mass of water condensed cond i

h

= 3.aturation enthalpy of steam at the partial pressure sat ot' vapor in the containment h,11 = enthalpy of liquid at wall surface temperature y

h = heat transfer rate during ti=e interval m affective heat transfer surface - area multiplier A

=

I = average temperature of heat sink surface during ti=e interval f = average bulk temperature of fluid adjacent to boundary B

during time interval.

h = heat transfer coefficient f.

Provide a table of the peak values'of containment atmosphere temperature and pressure for the spectrum of break areas and power levels analyzed.

RESPONSE

The containment atmosphere temperature and pressure responses for the spectrum of breaks are listed in the table below. Only one feedwater line break was analyzed since the break of a smaller feedwater line would result in pressurization less severe than that of the main feed-water line.

In all cases, the reactor was assumed to be operating at 102 percent rated power prior to the accident, 3

D-B Peak Pressure @ Time Peakfemperature@ Time Break Type (psig)

(sec)

( F)

(sec) 18-in. feedwater line 12.9 106.0 243.0 50.0 2

2.6 ft steam line 19.5 29.5 266.5 30.0 3.1 ft steam line 19.7 28.5 262.1 30.0 4.4 ft steam line 20.4 28.0 282.7 30.0 Double-ended steam line 21.4 27.0 288.1 25.0 g.

For the case which results in the marinnnn containment atmosphere temperature, graphically show the containment atmosphere temper-ature, the containment liner temperature, and the containment concrete temperature as a function of time.

Compa a the calculated containment atmosphere temperature response to the temperature profile used in the environmental qualification program for those safety-related instruments and mechanical components needed to

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i mitigate the consequences of the assumed main steam line break j

and effect safety reactor shutdown.

RESPONSE

i The case which results in the maximum containment atmosphere temperature is the double-ended main steam line break. The containment atmosphere, steel containment, and containment interior' concrete temperatures are plotted as functions of time on Figure 2.

FSAR Figure 15.4.4-4 also shows the con-i tainment vapor temperature and worst heat sink temperature as functions of

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time for this break. For a further discussion of the environrental quali-fication of safety-related instruments and components, refer to the response to part j.

h.

For the case which results zu maximum containment atmosphere pressure, graphically show the containment pressure as a function of time.

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RESPONSE

The case resulting in the maximum ccntainment atmosphere pressure is the double-ended main steam line break. A plot of atmosphere pressure vs time is shown on Figure 3.

1.

For the cases which result in the maximum containment atmosphere pressure and temperature, provide the mass and energy release data in tabular form.

RESPONSE

The mass sad energy release data for the main steam line break in the containment is listed in Table 1 below for the double-ended main steam line break.

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D-B Table 1

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Doubic Ended Main Steam Line Break Mass and Energy Release Data Time (sec)

Mass Flow (b/hr)

Energy Flow (Btu /sec) 0.00 0.0 0.0 l

0.01 5.997 E7 1.991 E7 i

0.05 5.885 E7 1.953 E7 0.09 1.044 E8 2.352 E7 O.14 1.326 E8 2.582 E7 0.18 1.725 E8 2.952 E7 0.22 1.798 E8 2.987 E7 0.26 1.819 E8 2.996 E7 0.30 1.822 E8 2.996 E7 0.35 1.815 E8 2.980 E7 l

l 0.40 1.804 E8 2.967 E7 0.50 1.773 E8 2.925 E7 0.60 1.738 E8 2.882 E7 0.80 1.650 EP 2.791 E7 1.00 1.571 E8 2.658 E7 1.20 1.472 E8 2.547 E7 1.60 1.265 E8 2.280 E7 2.00 1.041 E8 1.992 E7 2.65 3.971 E7 9.067 E6 3.25 2.737 E7 7.291 E6 4.00 1.927 E7 5.979 E6 6.00 1.461 E7 4.890 E6 10.00 1.258 E7 4.207 E6 13.28 1.186 E7 3.966 E6 16.00 1.047 E7 3.496 E6 22.00 6.190 E6 2.055 E6 28.00 4.800 ES 1.555 ES 30.00 4.000 E5 1.296 E5 Up to Aux. Feedwater 4.000 E5 1.296 E3 Isolation j.

For the instrumentacion and equirment located inside the contain-ment and required to (1) detect the steam line break; (2) initiate safety systems and (3) monitor the course of the accident, provide the following:

(1) A description of the tests which were/or will be performed to show that this instrumentation and equipment are/or will be qualified to perform their function before, during and after the accident.

Include the spectrum of environmental condi-tions for which tests were/will be performed and state the acceptance criteria. The instrumentation and equipment to be considered includes, but is not limited to the following:

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D-B (a) pressurizer pressure and level sensors and transmitters; I

(b) steam generator pressure and level sensors and transmitters; (c) main steam line pressure, differential pressure and flow sensors and transmitters; (d) primary system hot leg and cold leg temperature sensors and transmitters; (e) primary syatem pressure sensors and transmitters; (g) feedwater flow sencors and transmitters; (h) containment pressure sensors and trans-mictors; (1) valve operators and position switches; (j) elec-trical cables, motors and penetrations; (k) containment coolers.

Also identify any additional instruments and equipment required.

(2) A description of the separation and independecce.between redundant-sensors, cables and other equipment associcted with each steam generator and steam line.

(3) A description of the independence and separation between each steam generator and between each steam line.

RESPONSE

(1)

To assure that the instrumentation and equipnent are qualified to perform their function, the following information is provided:

Pressurizer level transmitters.

RC pressurizer level (0-320 in. H.,0) type tested to 300 F and 60 psig -

allowable deviation i 10 percent Eype test data error i 3.6 percent span shif ted i 3 percent after the test.

FSAR subsection 3.11.2.1.1 discusses the qualification of this instrumentation.

Steam generator level transmitters (NSSS supplied)

Steam generator start up range level (0-250 in. H 0) type tested to 300 F and 60 psig - allowable deviation i 10 percent.

e test data error i 3.6 percent span shifted i 3 percent af ter test. FSAR subsection 3.11.2.1.1 discusses the qualification of this instrunentation.

Steam generator level transmitters (Other than NSSS supplied) used as input to SFRCS FSAR subsection 3.11.2.2.3 discusses the qualification of this instru-mentation. In addition, the environmental test temperature profile for these Rosemount 1152 transmitters is shown on Figure 4.

Main steam line pressure switches These pressure switches are not located in the containment. They are qualified as discussed in FSAR subsection 3.11.2.2.9.

Reactor coolant outlet temperature RTD Reactor System Outlet temperature (520-620 F). This string was not type te:;.ced for loss of coolant accident. The resistance temperature 1

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FB detector (RTD) is supplied with a connector head that has been pressure tested to 60 psig. The continuous rating of the connector head material is 300 F.)

PSAR section 3.11.2.1.1 discusses the qualification of this instrumentation.

Reactor coolant wide range pressure tramsmitters and narrow range pressure transmitters (input to RPS - Rosemount 1152)

FSAR subsection 3.11.2.1.1 discusses the qualification of this instrumentation.

See Figure 4 for temperature profile for the Rosemount 1152 environmental test. RC loop pressure (0-2500 psig) i type tested to 318 F and 90 psig - allowable deviation i 12 percent during; i 4.0 percent after. The test data indicated i 9 percent during and less than i 2 percent after the accident. The lesser temperatures of the steam line break of 288 F would produce even less error.

Steam generator outlet pressure transmitters Steam generator outlet pressure (0-1200 psig) type tested to 318 F and 90 psig - allowable deviation i 10 percent. Type test data error i 3.6 percent span.

Span shift i 3 percent after test. The qualifica-tion is described in Topical Report BAW - 10003A.

Reactor coolant flow transmitters FSAR subsection 3.11.2.1.1 discusses the qualification of this instru-mentation.

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Feedwater-steam genernror differential pressure switches These differential pressure switches are not located in the containment.

They are qualified as discussed in FSAR subsection 3.11.2.2.8.

Containment vessel pressure transmitters These pressure transmitters are not located in the contaimeent. They are qualified as discussed in FSAR subsection 3.11.2.2.4.

Valve operators (containment isolation valves)

Valve operators have been qualified by a generic test performed in accordance with IEEE-382-1972 requirements. The tested operators were periodically cycled under simulated loads in a steam environment for 30 days. Environmental conditions reached 105 psig at 340 F as shown on Figure 5 and were step decreased during the first 4 days to 10 psig at 200 F where it was maintained for the remaining 26 days. The operator was shown to be acceptable due to the ability to cycle periodically under the test conditions.

Electrical cables In addition to the qualification discussed in FSAR subsection 3.11.2.1.5, the following tests were performed, as shown in Figure 6.

For BIW cable, l

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I the cable was subjected to;53 psig at 300 F for 15 minutes. Then the 1

temperature was reduced to 252 F for 10 days at 16 peig. For Kenite cable, the cable was subjected to 82 psig at 320 F for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Then the temperature was reduced to 228 F for 7 days at 5 psig.

The acceptance criteria for the BIW cable was the maintenance of 600 V during the test. For the Kenite cable, the acceptance criteria was that 600 V be maintained between adjacent conductors as well as the mainten-ance of 50 amps / conductor for No. 6 and 12 amps / conductor for No. 12 Conductor.

Electrical penetrations Iri addition to the qualification discussed in FSAR subsection 3.11.2.1.6, the following tests were performed, as shown in Figure 7.

Penetration was installed in the test chamber. Using steam the chamber was allowed to reach normal operating temperature (100 F - 125 F). The temperature was raised to 300 F at 55 psig for 15 minutes. Then the chamber rammined at 255 F and 20 psig for 23 3/4 hours.

Before and after the teges, acceptance was indicated by leak testing within the range of 10 STD cc/sec. Acceptable continuity, dielectric, and insulation resistance was also required before and after testing.

Containment air coolers FSAR subsection 3.11.2.1.3 discusses the containment air coolers.

Qualification of the equipment is based on an environmental test per-formed by the vendor where a similar fan motor was started and operated in a steam environment at 60 1 5 psig and 287 F for two hours. The performance of the test motor is used to qualify the Davis-Besse Unit 1 equipment based on a certified calculation relating motor and fan design and performance characteristics.

During the entire test period, readings of power, temperature, pressure and resistance of RTD's were taken. After completion, Megger tests were run. The motor was then restarted and operated for one hour.

Megger readings of the insulation resistance were again taken. The motor was then disassembled and checked for evidence of coating failure.

The motor was then assembled into a fan unit and a low-speed heat run was performed, with stable RTD readings being observed for one hour.

The acceptance criteria for the above instrumentation was that the instrument perform its function after the testing within the acceptable instrument accuracy.

(2) A discussion of the separation and independence between redundant sensors, cables, and other equipment is provided in FSAR section 7.1.23 and in response to NRC question 8.1.2 (3) The independence and separation between each steam generator and between each steam line is shown in FSAR Figures 1-9, 5-4, and 3-35.

Figures 3-36 and 3-37 show the routing and restraint system of the steam lines from each steam generator.

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DAVIS-BESSE NUCLEAR POWER STATION DOUBLE-ENDED MAIN STEAM LINE BREAK HEAT TR ANSFER COEFFICIENT VS TIME FIGURE 1

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- CONTAINMENT VAPOR MAX TEMPERATURE 288.11 F

@ 25.0 SEC. (INCLUDED IN FSAR FIG. ISAAAl 200--

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DAVIS-BESSE NUCLEAR POWER STATION DOUBLE-ENDED MAIN STEAM LINE BREAK TEMPERATURE VS TIME FIGURE 2

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DAVIS-BESSE NUCLEAR POWER STATION DOUBLE-ENDED MAIN STEAM LINE BREAK CONTAINMENT PRESSURE VS TIME FIGURE 3

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ACTUAL DATA

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THEORETICAL DATA START 350 F l

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TIME FROM START OF TEST DAVISrBESSE NUCLEAR h0WER STATION ROSEMONT 1152 TRANSMITTER ENVIRONMENTAL TEMPERATURE TEST PROFILE FIGURE 4

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TEMPERATURE INC. TO 340 F F 200 IN LESS TilAN 4 MINUTES w

19 SEC RISE TO 334 F/110 PSIG.

TEMPER ATURE INC. TO 341 F IN LESS TilAN 3 MINUTES.

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+ 6 HRS TEST DAVIS-BESSE NUCLEAR POWER STATION CONTAINMENT ISOLATION VALVE MOTOR OPERATORS ENVIRONMENTAL TEMPER ATURE TEST PROFILE FIGURE 5

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DAVIS BESSE NUCLEAR POWER STATION ELECTRICAL CABLES ENVIRONMENTAL TEMPER ATURE TEST PROFILE FIGURE 6 e

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TIME DAVIS-BESSE NUCLEAR POWER STATION ELECTRICAL PENETH ATIONS ENVIRONMENTAL TEMPERATURE TEST PROFILE FIGURE 7 I

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