ML20205B344

Forwards Response to NRC 870303 Second Request for Addl Info Re Firewater Cooldown from 82% of Full Power (Safe Shutdown Cooling)

ML20205B344
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/20/1987
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
P-87110, TAC-63576, NUDOCS 8703270593
Download: ML20205B344 (16)

Text

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2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 March 20, 1987 Fort St. Vrain Unit No. 1 P-87110 U. S. Nuclear. Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Additional Information for Analysis of Firewater Cooldown for 82% Power Operation

REFERENCE:

1) NRC letter Heitner to Williams, dated February 3, 1987(G-87031)

2) PSC letter Warembourg to Berkow, dated December 30, 1986(P-86683)

3) PSC letter Williams to Berkow, dated January 15, 1987 (P-87002)

4) PSC letter Brey to Berkow, dated February 17, 1987 (P-87055)

5) NRC letter Heitner to Williams, dated March 3, 1987(G-87060)

Dear Mr. Berkow:

Reference 1 requested that PSC provide additional information concerning the firewater cooldown from 82% of full power in addition to that which was presented by PSC in References 2 and 3. PSC provided responses to NRC's first request for additional information in Reference 4.

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P-87110 Paga 2 March.20, 1987 Attachment 1 provides PSC's response to the NRC's second request for information presented in Reference 5. This additional information is provided to. enable the NRC to complete its review of PSC's analysis of a firewater cooldown from 82% of full power.

If you have any questions about the responses in Attachment 1, please contact Mr. M.'H. Holmes at (303) 480-6960.

Very truly_yours,

• W H. L. Brey, Manager -

Nuclear Licensing and Fuels Division HLB /AHW:jmt Attachment cc: Regional Administrator, Region IV Attn: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain

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,.  ; NRC-SECOND~ REQUEST FORl ADDITIONAL INFORMATION

• CONCERNING SAFE SHUTDOWN COOLING-FORT.ST. VRAIN

1. -DOCKET NO. 50-267:

NRC QUESTION 1 I . What is'~the status of.the hot and cold reheat header isolation. valves-during a firewater-cooldown?

PSC-RESPONSE'1 Firewater cooldown is postulated.following a High-Energy Line Break

's (HELB), an Appendix R fire, an Earthquake, or a Tornado. After.

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- firewater- has-been aligned for cooldown and subcooled firewater flow , 5 L ' has been established at the outlet of a steam generator, _ forced circulation Jof helium through the reactor core is. recovered with one a circulatorf driven -by. firewater supplied to its water turbine.

Whenever reheat- . steam is not used for driving a helium circulators' l steam. turbine, its . reheat steam ~ (inlet and outlet) isolation -valves.

are normally closed, per System Operating Procedure S0P 21-01. .In -

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, the event of a loop shutdown, the ' circulators' cold reheat steam

. header . isolation valves,; bypass block- valve and the hot reheat steam

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~stop check' valve are automatically closed, effectively isolating that-loop's reheater,

!- Actuation of SLRDIS' following a HELB would.close the reheat steam isolation valves on both loops. Manual action by the operators is

required to reopen the valves, but the operators would keep these L - valves closed during a-firewater cooldown since they are not in the firewater flow pa th . .. If the HELB were'in the reheat steam piping, l the operators would prevent venting the remaining reheat steam

' inventory into the Reactor Building through the ruptured pipe by

< -- manually closing the reheat steam isolation valves or assuring that l they remain closed after being closed automatically by SLRDIS.

[ - For an Appendix R fire, seismic event, or tornado the status of the valves would (without operator action) be either open or closed

- depending. on the severity or location of the accident and the resulting automatic-actions of the plant control systems.

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-NRC-QUESTION 2:

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Are'thevalvesditherclosedoroperableforautomaticclosureshouldo l La leak develop in~ the reheater

tubes?-

i PSC RESPONSE'2 _

Asi discusse'd in PSC Response ~ 5, it has been determined th'at the.

reheaterJtubes' will not rupture in the event of a firewater cooldown

. . following _ a ' 90.. minute . interruption of forcedJ circulation after

. freactor operation at 105% power. s

<If .aJ leak occurs in the reheater tubes of:a steam generator during-

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normal operation, the primary coolant helium in the.PCRV . would . leak into the ' reheat steam piping-system, which is at a pressure lower

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-than -the; primary- coolant pressure. Radioactive gases and particu.lates~ within the ' primary coolant > could then pass =into the (J

s . - reheat steam. piping system andtthen.out to the environment. via -the main condenser.-

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. IR'adiation : detectors monitor. the fluid inside the loop hot reheat- -

3; steam headers::forl indications of. reheater tube ruptures.. Upon detection .of radioactivity above the trip setpoint.in the hot reheat piping,-a signal is sent.via the PPS system to automatically close- -

the reheat _ steam.' isolation valves and the feedwater and main' steam header isolation val.ves, thus completely' isolating the affected loopi The - reheat - steam isolation valves are automatically closed upon ,

actuation of thel SLRDIS following 'a HELB.- The SLRDIS- Recovery Procedure OPOP XII, Checklist Step 2, requires the. plant operators to L .

actuate'all valve hand: switches closed, which were closed-~by- SLRDIS actuation, to assure -they do not reopen when SLRDIS is reset.

Following a' HELB in the Turbine Building the SLRDIS recovery-

! -procedure, 0 POP-XII, Checklist Step 7, requires the plant operators to remove the six hot reheat activity monitor modules, in order to g a .. prevent postulated malfunction of these non-environmentally qualified monitors from causing a loop shutdown, including shutdown of the :

recovered':(operating EES)' loop and thus interferring with SLRDIS recovery.

The reheat steam isolation valves and the PPS hot reheat steam header radiation detectors are seismically qualified and would therefore

-remain operable following an earthquake or tornado. The reheat steam V'

isolation valves would be closed by the radiation detectors in the

• < event of a reheat tube break.

i: ' For- a: fire, the worst condition that could result is inoperable

' reheat steam isolation valves and hot reheat steam header radiation

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detectors. That is, the reheat steam header isolation valves could be assumed to be open. This condition is considered acceptable because the acceptance criteria delineated in PSC Letter Lee to Johnson, dated August 17, 1984 (P-84281), for an Appendix R fire

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v-outside the Congested Cable Areas:(CCA). include:-

. 21. "For any: single fire-in 'a non-congested cable area means shall be available to: shut'down and cool.down the reactor in a manner.such thath ._no fuel; damage. occurs . :(i .e. maximum : . fuel particle

~ temperature does not exceed 2900 degrees F). There.shall be 'no simultaneous _ rupture ( of both a' primary: coolant boundary.and the f; associated = secondary. containment. boundary.. such that :no

~ unmonitored radiological releases of primary coolant occur."

-2. Thel specific: criteria for. single failures in conjunction with a fire is:-

" Systems: used- to ensure the post fire safe . reactor

-shutdown /cooldown capability need not be designed to meet seismic Category -I : criteria,- single failure criteria, or other.. design basis accident criteria, except where required.for other-reasons,-

e.g., because of -interface with or impact on existing safety systems, or because of adverse valve actions due'to fire damage."- -

Asf discussed .in PSC Response 5, the reheater. tubes will not rupture

-in the event of a cooldown following a fire outside the CCAs; Even though the- secondary. containment boundary could be postulated to be

. breeched-(open reheat steam isolation valves) following.a fire, the-primary coolant boundary-would remain intact, meeting the acceptance-criteria for a fire outside the CCAs.

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Attachment to-P-87110 Page 4 NRC QUESTION 3-Are-there initiating events for which this system is not. qualified or does-not meet regulatory requirements, such as an earthquake .or

-single fire?

-PSC RESPONSE 3 The radiation detectors and isolation valves described in PSC~

-Response 2 are seismically qualified. The radiation detectors are not environmentally qualified while the reheat steam isolation valves are environmentally qualified and would .be available for manual isolation in the event of a HELB, if SLRDIS has not already closed these valves. These instruments and valves are not relied upon as part~.of the ~ Appendix R . fire protection model. As discussed in PSC Response 5, it has been demonstrated that the PCRV pressure boundary is maintained during firewater cooldown and that the reheater tubes are not predicted to rupture following a HELB, an Appendix R fire, seismic event, or tornado.

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NRC QUESTION 4 .

What .safetyE benefits .would come -from closing 'the valves before

restoring helium circulation? '

PSC RESPONSE'4 Closing the: reheat = steam isolation valves 'following an accident-requiring the alignment.of a firewater- cooldown, if not already closed; by.' automatic action, would. in' the event'of a subsequent

. - reheater tube rupture result'in containing the. primary coolant helium i ;in the'PCRV since the closed reheat steam isolation valves would form a secondary pressure boundary.

Closing the reheat steam isolation valves would establish 'two.

redundant barriers to prevent release of primary coolant._ Reheater tube ruptures during the cooldown with' firewater were evaluated and are- not expected to occur (see PSC Response 5). Because'of the many-operator actions occurring during the first few hours of cooldown on firewater, requiring Lthe . operators to close the reheati steam

. isolation valves during .this period ~might' introduce unnecessary complications.

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LNRC QUESTION 5'

-What? sa'fety benefits. would come from.being able to pressurize the reheater with dry gas'before restoring helium circulation?-

PSC RESPONSE 5 Reversing' the: pressure differential 'across the reheater tube wall by

. pressurizing the reheater with dry gas would change' tube stresses from compressive to tensile and therefore eliminate the probability.

of tube buckling. But the analysis of the event has'shown that the- s

<- likelihood.of the reheater tube buckling is practically non-existant.

The reheater tube buckling ' analysis,- reported in GA~ Technologies 3

Report GA-909204,~- Issue N/C,.and submitted to-the-NRC as Attachment'7' ,

of PSC-letter.P-87055, utilized a maximum temperature of 1660 degrees

F 'with a maximum . pressure differential of 350 psi- for a period of approximately two hours. This_ analysis concluded that'the potential

for tube'. buckling arises when these temperature and pressure

' conditions are maintained for more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. This ' analysis is conservative since the: model assumes tube buckling occurs'when a single element of.the tube reaches yield stress. The 1660 degrees: F.

and 350 psi. conditions- are representative of a 105 percent power

. case. The calculated maximum tube temperature at the 87.5 percent

• power . level is reduced to 1501 degrees F. From Table 4.2 of the GA-909204 r? port, the time for the reheater tube to reach: buckling at 1500 . degrees F using the maximum external-pressure used for the

~

. buckling calculation of 832 psizis 196 hours0.00227 days <br />0.0544 hours <br />3.240741e-4 weeks <br />7.4578e-5 months <br />. As - the . temperature of 1500 degrees F is not expected to be exceeded for more than two hours-and the external pressure is expected to be less than 700 psi at any  ;

time 'during the event, there is a significant margin of safety. It r

is noted that for the Appendix R cases (83.2% power level) reported -

in GA Technologies Report GA-909269, Issue A, this. temperature has -

been calculated to be slightly less than 1500 degrees F.

The' added complication of introducing the pressurized dry gas to the reheater seems to be an unnecessary precaution. Aside from any complications that a dry gas pressurizing system might introduce to

-the many actions that are already- required of the operators, the-design and ' operation of such a system may introduce more problems than the benefits it might provide. For instance, overpressurization

,' of a hotter than average reheater tube by the dry gas system could -

-result ~fn rupture of this tube.

l 4

Attachment to P-87110 Page 7 NRC QUESTION 6 Page 14 of Attachment 1 to P-87055 refers to 'the results of special testing of simulated firewater' supply to the steam generators. and circulator Pelton drives. The text of Unusual Event Report 76/05A indicates that Test T-30 yielded only 540 gpm supply to the steam generator. What was the circulator speed achieved during this test?

What were the steam generator supply flow and circulator speed during Test T-30A? Test RT-403 achieved circulator speeds in excess of 700 RPM on each circulator. What was the steam generator supply flow during RT-403? What circulator speeds and concurrent steam generator flow rates were obtained during any testing after installation of the j emergency water booster pumps? What were the simulated firewater L supply pressure (booster pump suction), steam generator back l pressure, and Pelton wheel supply pressure during each of these tests for which data are provided?

FSC RESPONSE 6 i

Table 6-1 is a summary of the T-30 test series. Figure 6-1 is the T-30 test flow schematic. In the initial T-30 test "A" circulator only was tested and this was for a low steam generator back pressure /high flow test case. The recorded circulator speed was 636 rpm which did not satisfy the approximate 700 rpm required speed for cold core conditions. Instrumentation upgrades were made and additional testing identified as T-30A were conducted. These are also summarized in Table 6-1. "A" circulator speed remained low for the one test while "B" circulator speed was observed to improve with subsequent tests. It was later concluded based upon RT-403 testing that the panel and data logger instrumentation used for T-30 speed readouts were significantly in error. RT-403 used the speed / wobble scope for circulator speed measurements. '

The T-30 series tests identified excessive pressure drop in valves HV-21257 through HV-21260 which are located in the flow path to the water turbine drives. These valves were subsequently modified to decrease pressure losses. Testing conducted in RT-403 evaluated circulator performance on water turbine drive on "A" and "B" circulators before the valve modifications and on all circulators after the valve modifications. The results of RT-403 testing are summarized in Table 6-2. The RT-403 flow schematic is shown in Figure 6-2. The test indicated that the valve pressure losses were i decreased by almost 50 psi down to 5-7 psi and that circulator speeds significantly increased satisfying the FSAR criteria at that time. ,

The criteria for the test was to simulate firewater cooling by I throttling the condensate pressure to maintain 113-115 psig at PI I 31204, which senses the pressure upstream of the Pelton wheel. The steam generator supply flow and other users of condensate and their flow rates during the RT-403 testing were not recorded.

The emergency water booster pumps were installed in 1979 to enhance the performance of the circulators when being driven by the water turbine drive on firewater. The primary requirement of the emergency

l' .

Attachment

-to P-87110-LPage 8

- water booster. pumps was'to assure a circulator performance when_being driven..by the water turbine drive .at least equivalent to that attainable with 175 psid.at the water turbine nozzle. -Typically this:

1-will provide. a primary coolant helium flow varying between 3 to 4%

under hot-- core ' accident - conditions.

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emergency . . water booster- pumps .all~ testing. of the water turbine-

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Ldrives is performed to the requirements of Technical Specification SR 5.2.7.a. See. response 'to NRC-Question.7 for discussion of testing ~

. performed since the . emergency water.-booster pumps have been installed. .

.Upon review of the test data from Tests T-30 and.T-30A, PSC concludes that while the dataiis generally in' the range .to support recent analyses,' it: is not valid to rely on-the data for direct support of' these analyses cInstead, PSC proposes that a special test be run to-provide accurate and sufficient data, as described below.

'The tests T 30 and T-30A were conducted so'long ago that the results do not warrant current detailed evaluation. Since the time of these.

tests, several valves (HV-21257 through HV-21260).have been modified-to decrease excessive'-pressure drops. . Addltional.ly, the Emergency _

Water Booster Pumps .have been installed and then enlarged, and the.

new 6" vents-have been installed to allow: increased flow through 'the steam generators.

Therefore,- PSC considers a more prudent review path to be the running of.a special test of simulated firewater. to a circulator Pelton-drive. This test' is discussed in PSC Response 7. 'Any testing of firewater flow rates through the steam generators would still be an

, approximate simulation requiring significant analysis to correlate to actual conditions. . Therefore, PSC would still rely on the existing analysis to' verify those flow rates.

7

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' Concurrent Circulator Speed and Steam Generator Testing Performed in T-30. Test. Series 1

Test T-30'6/2/76(2) T-30A6/16/76:(3)

Test Data PV-22129 Closed PV-22129 Closed. PV-22129 Open PV-2229 Open PV-2229 Open PV-2229 Open- '

Test Point Number 1 2 3 4 1 2 3 4 5 6 7- 8 Steam Generator 770 770 270 640 640- 640 640 690 700 --

'780. --

Loop Flow,(l) 103 lbm/hr ,

Steam Generator -- -- -- -- 1210 1235 1235 1320 1340 1360 1425- 1500 Loop Flow, gpm(l)

Emergency Cond. 115- 125 305 290 125 125 113 115 122 137 115 .'135 Header Pressure at PI 31204.

Elv. 4819, psig Steam Generator 160 75 75 0-25 58 60 60 64 63 50 51- --

Exhaust Pressure at PI-22129, psig Circulator Pelton -- -- -- 136 137. 138. 139 151 161 140 1501 160 Nozzle Supply Pressure at Test Gauge Elv. 4770, psig (same as Test Gauge 2 in RT-403 Testing)

Circulator -- -- --

636 0' 600 625' 666 750 731 '750 790 Speed, rpm "A" "A" "B" "B" "B"- "B" "B" "B" Notes (1) Two measurements available for T-30A '

(2) Helium temperature was 1030F and helium pressure'286 psia (3) Helium temperature was 1280F and helium pressure 334 psia-

Table 6-2 .

RT-403 Simulated. Firewater Circulator Speed Testing Prior HV21257 through Test HV21260 Valve Modifications (l) AfterValveModifica4jQns 10/3/76-and 10/5/76', 1 9/28/76 Test Gauge 1 114 115 114 114. 114' 114-psig (same location as PI 31204 in T-30 series tests)

Test Gauge 2, 136 136 133 131.5 -- --

psig Test Gauge 3, .78 81 128 124.5 -- --

psig Test Gauge 4, 74 86 130 124.5 -. --

1 psig j Circulator Speed, 630 650 760 790 790 ,770 rpm from "A" .

"B" "A" "B" "C" "D" speed scope Notes

1. Helium temperature was ll70F and helium pressure was 305 psia.

2. Helium temperature was 1190F and helium pressure was 338 psia for testing of "A" circulator..

Helium temperature was 1220F and helium pressure was 333 psia for testing of "B", "C", and "D" circulators.

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<NRC QUESTION 7 SR 5.2.7.2' is' indicated to..be -performed with 115 psig suction pressure at'the emergency booster pump from throttled- condensate.-.

' Provide .the results Lof SR 5.2.7.2 in -tems of booster pump suction pressure,L -booster: pump differential ' pressure, circulator-.. speed, simulated -firewater flow rates from the booster pump and to the'

. Pelton wheel, .and primary system flow at rated helium. inventory.- Has

-SR: 5.'2.7.2 been' performed in conjunction with simulated firewater flow to the steam generator? If so, provide ~the above results plus the steam generator flow and back pressure measurements.

PSC RESPONSE 7 SR_- 5.2.7.2 should: be SR 5.2.7.a. -Surveillance test SR 5.2.7.a has been replaced.with SR 5.2.7.a2-A effectiv'e November 30, 1984, after.

the Emergency Water Booster -Pumps .were upgraded. The following.

results are taken from SR 5.2.7.a2-A, conducted July 18, 1985:

WATER TURBINE USING CONDENSATE AT REDUCED PRESSURE TO SIMULATE

-FIRE PUMP PRESSURE ON LOOP I CIRCULATOR-P-2109 Booster Pump Suction Pressure - . . . . . . . . . . . . . . . . . . y. . . . . . . . . . . 115 psig Ci rcu l a to r' S peed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2100 rpm -

P-2110 Booster Pump-Suction Pressure .............................. 115 psig Circulator Speed............................................ 2200 rpm

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WATER-TURBINE USING CONDENSATE AT REDUCED PRESSURE TO SIMULATE FIRE PUMP PRESSURE-0N LOOP II CIRCULATOR' ,

P-2109:

' Booster Pump Suction Pressure .............................. 113 psig Circulator Speed............................................ 2100 rpm P-2110

, Booster Pump Suction Pressure .............................. 114 psig i Circulator Speed............................................ 2000 rpm This surveillance test procedure does not require measurement of the Booster Pump differential pressure, the flow rate of the simulated firewater to the Pelton wheel, or primary system flow rate. During the performance of this test, the circulator inlet helium temperature i was 120 degrees F, and the reactor pressure was 64.3 psia. From this

a helium density.of .04 lb/ cubic foot was calculated. A minimum

. acceptable circulator speed of 1825 rpm was determined based on this l calculated density, and the circulator being driven by the water

! turbine drive with 175 psid across the water turbine nozzle. During

[ the test, the circulator speed was increased until an acceptable l speed above this minimum speed was attained. This is the speed noted l above.

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L l

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Attachment to P-87110 Page.10 SR 5.2.7a2-A has not'.been performed in conjunction with simulated firewater flow to the steam generators. Further, if this test had been performed the results would not be applicable -to the new firewater flow paths.

While SRS.2.7a2-A measures helium circulator speed with the simulated.

firewater supplied via the booster pumps, the helium flow rates and simulated firewater flow rates are not routinely determined.

Calculations were recently performed to determine the helium and simulated firewater flow rates involved in the July 18, 1985 test of the circulators. These calculated flows were within the range of the P2 shutdown analysis. However, the only way to verify the actual flow rates is to perform a special test, similar to SR 5.2.7.a2-A, with an increased amount of data being recorded. PSC will perform this test within the next few ' weeks and submit this additional information to the NRC in answer.to this question.

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