ML19260A765

Forwards Info Re Potential for Steam Generator Water Hammer, in Response to NRC 790911 Request

ML19260A765
Person / Time
Site:	Fort Calhoun
Issue date:	11/12/1979
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 7912030212
Download: ML19260A765 (6)

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Omaha Public Power District -

1623 HARNEY a OMAHA. NESRASKA 68102 e TELEPHONE S36 4000 AREA CODE 402 November 12, 1979 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid , Chief Operating Reactors Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

Docket No. 50-285 Gentlemen:

The Omaha Public Power District received a letter from the Commission, dated September 11, 1979, requesting that information be provided in regard to the potential for steam generator water hammer at the Fort Calhoun Station. In response, the attached information is provided.

Sincerely,

.A i

, 4 1% "

1 g W. C. Jones Division Manager y Production Operations WCJ/KJM/BJH:Jmm ec: LeBosuf, Lamb, Leiby & hacRae 1333 New Hempshire Avenue, N. W.

Washincton, '. C. 20036 1460 348 y

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79n030

Request 1 Provide information that demonstrates that the feedwater system aa:4 steam generator water 'evel at your facility have been subjected to those transient conditions that are conducive to water hammer, i.e., the addition of cold feedwater or auxiliary feedwater to steam-filled feedwater piping and feed-ring. See NUREG 0291 page 4 that was forwarded to you on September 2,1977.

Include the following:

1.1 Describe the expected behavior of steam generator water level as a result of reactor trip from power levels greater than 30% of full power. Include actual plant measurements of steam generator level

.nnd other available related data such as feedwater flow and auxiliary feedwater flow.

Response

• * ** Time Feedwater Steam Iowest Required Flow Rate Power Ievel Gen. Level Steam Gen. for Remvery at at Trip at Trip level after of Pre-Trip Trip Date (%) (%) Trip (%) Ievel (hrs.) (x106 lbs hr 1)

(1) 5-15-75 31 68 50 1.3 0.7 (2) 2-12-76 86 66 30 0.5 2.6 1.7 2.1 (3) 2-21-76 71 69 21 (4) 5-28-76 80 70 35 1.0 2.4 (5) 8-21-79 100 70 27 1.3 3.1

• 2 e upper 1" level indication tap is used as a reference point 111 defining the 0-100% steam generator water level. A distance 37.4" lower than the tap is defined as the 100% level and a distance 169.9" lower than the tap is taken as the 0% level.

me trip set point is at the 31.2% level, which is situated 58" below the normal water level.

• • It should be noted that this tine for recovery is operator dependent.

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CausesofTrg

{l) Dropped CEA (CEA No. 25 Group 2) .

(2) Spurious actuation of Thermal Margin Ioa Pressure Trip Channels C and D.

Trip due to induced signals on the temperature instrumentation loops.

(3) Ioss of 161 KV supply to the house service buses, tused by failure of fast transfer due to relay.

(4) Off site grid load rejection.

(5) Ioss of D.C. power to E.H.C. unit.

1.2 Provide the number and causes of loss of feedwater events during the operational history of the plant. You may refer to material submitted previous 1y.

Besponse here have been threr loss of unin feetwater events during the plant's history:

Item Date Cause (i) 3-07-74 Scheduled cmplete loss of off-site A.C. power during trip test; main feedwater punp trip.

(ii) 2-21-76 Ioss of 161 KV supply to house service buses followed by failure to fast transfer; main feedwater pump trip.

(iii) 8-22-77 Rxtentary loss of 161 KV supply to house service buses followed by failure to fast transfer; main feedwater pump trip.

It is emphasized that the auxiliat'f feedwater systen was available to provide water to the steam generators during each of these events.

1.3 Provide the number and causes of loss of off-site power events during the operational history of t):e plant.

Rest:ense m ere have been three loss of off-site power events as shown in our response to itms 1.2 (i), (ii) , and (iii) .

Request 2 If administrative controls have been adopted to limit the flow of auxiliary feedwater for the purpose of reducing the probability of water hammer, show when they were adopted and give the answers to items 1.1,1.2, and 1.3 for before and after such controls were established.

Response

No administrative controls have been adopted to limit the flow rate of the auxiliary feedwater.

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Request 3 .

If administrative controls have been adopted to limit the flow of auxiliary feedwater for the purpose of reducing the probability of water hammer, show that an adequate water inventory and flow will be maintained to accom-modate all postalated transient and accident conditions.

Restonse

'Ihe question is not anplicable (as per request .2.)

Request 4 If auxiliary feedwatar flow in your facility is not at present initiated auto-matically for normal and accident events, present your evaluation of whether automating the actuation of auxiliary feedwater might increase the probability of inducing steam generator water hammer. One of the signals that woald automatically initiate the flow of auxiliary feedwater would be the steam generator low water level. This set point should be above the top of the main feedgater sparger to reduce the probability of steam generator water hammer.

Response

At the Ebrt Calhoun Station there are two ways in which the auxiliary feed-water can be supplied, namely, either:

(a) Through the auxiliary feedwater line, using the auxiliary feedwater nozzle.

or (b) 'Ihrough the auxillary feedwater line, using the main feedwater ring.

At present, method (b) is pre erred and used during normal heatup and cool-down. However, it is anticipated that when the activation of auxiliary feedwater flow is automated, then route (a) will be etployed.

An investigation into the possibility of water hamner at the Fort Calhoun Station

• has been presented to the Ccmnission. The investigation found that the Fort Calhoun Station had the shortest practicable horizontal pipe run adjacent to the steam generator, namely 29", which included the feedwater nozzle and the tee connection to the main feedwater ring. This piping configuration is wwmmble to the other plants on which water ha:trer tests have been pes.fu1.M with no fluid flow insMhilities being encountered.

Furthermore, the location of the main feedwater ring is well below normal steam generator water levels, which reduces the probability of steam generator water level dropping sufficiently to uncover the main feedwater ring.

• Auxiliary Feedwater and Main Feedwater, Water Hamner Analysis for Fort Calhoun Staticn Unit 1, by Nuclear Services Corporation, January 22, 1976.

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Also, in the past when the steam generator level dropped to below the main feedwater ring (1.1. item = 2,3 and 5) no water ha:mer was observed. mis in itself mnfirms that the previously discussed factors of geonetry and water level, which are unique to the Fort Calhoun Station, have indeed negated the possibility of water hanner occurrence.

If, as anticipated in the future, the auxiliary feedwater nozzle will be used in the autcraatic mode to supply the aux 111ar'f feedwater, then two factors have to be considered *:

(i) 'Iae opening of the control valve ~:

It is expected that during the autcmatic node of auxiliary feedwater in ec-tion, the centrol and shutoff valves will open as the auxiliary feed-water pumps are starting. 'Ihus, no pressure dis-continuities will exist in the system and there will be no possibility of water hanmer occurring.

(ii) Condensation of steam between the check valve and the steam generator.

'Ihe. auxiliary feedwater line feeds the steam generator through an open ended pipe rather than through a sparger as is the case for the main feed line. 'Ihus ,

in thc. auxilia:n/ feedwater configuration there is no

~

supply of cold water downstream of the steam generator nozzle that can be drawn back into the feed line to create a cold water seal and hence a steam-water slugging phernrnon.

Should a slug of water from the steam generator enter the auxiliary feedwater nozzle when the auxiliary feedwater line is flowing par +1a11y full, it could not be accelerated back through the auxiliary feedwater nozzle for the following reasons. 'Ihe water in the steam generator at the water-steam interface is at saturated conditions. Any depressurization of steam in the nozzle would be transmitted back through the water slug causing it to flush to stean. Thus, the slug steam interface would be in a flashing rather than a condensing node preventing both rapid depressurization and the rapid acceleration of an inwessible slug of liquid.

It is therefore clearly evident that using the auxillary feedwater line and discharging eitler through the main feedwater ring or using autcraatic initiation and discharging through the auxiliary feedwater nozzle will not increase the likelihood of inducing steam generator water hanner.

"Auy11iary Feedwater and Main Feedwater, Water Hanmer Analysis for Fort calhoun Station, Udz 1, by Nuclear Services Corporation, January 22, 1976.

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Request S_

Describe the msons that sal' he used to ncnitor for the occurrence of steam generator water hanc ~ and possible damage from such an event.

nclude all instrumentation that will be employed. Describe the inspections that will be performed and give the frequency of such inspections.

Response

No formal nonitoring of steam generator water hanmer is performed at the Fort Calhoun Station, since as to date no water hammr proble:s have occurred in the main feedwater piping.

On February 27, 1976, the District subnitted to the Comission a report entitled "AMandum to Secondary System Fluid Flo.s Instability Peport -

Fort Calhoun Station Unit No. 1." 'Ihe aMandum provided results of a cxxtputer study done by Nuclear Services Corporation which analyzed the main feedwater and auxiliary feedwater systems. 'Ihe study considered the patential for the occurrence of a water hamner event and the effects on the piping system caused by either uncovering and draining of the feed-water sparger ring or by supplying auxiliary feedwater to the emergency feed nozzle. 'Ihe study misted of three parts; the first being a description and evaluation to detrzmine in what operational node water hamner is possible, the second consisted of a series of thermal hydraulic analyses to detarmha the conditions uncbr which a water ha"mer could occur and to quantify the magnitude of the water hanmer force, and the third part to deternura the response of the piping systems to the postulated water henmer events. 'Ihe analysis resulted in the conclusion that the probability of a water hammr occurrence is acceptably 1cw, as confirmed by plant operation experience.

It is expected that the plant staff would recognize water hamner by the sound generated, and possibly by coincident spiking on the secondary side pressure / level instrumentation. In addition, it is expected that any water hanmer would cause a significant discontinuity on the feedwater ficw recorder in the control room.

Finally, water hamuer could be recognized by lcng-tam equignent deteriora-tion such as displaced piping and broken snubbers which would be recognized by inservice inspections performed in accordance with the requirenents of the Fort Calhoun Technical Specificatic21s. Other signs would also include cracked or displaced insulaticrl.

Request 6 Describe the reporting procedures that will be used to document and report water hammer and damage to piping and piping support systems. Such reports wern requested in our letter to you dated September 2,1977

Response

If a significant water hanmer occurred affecting safety related piping, an IER (Iicensee Event Report) would be subnitted.

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