ML18219D053

D.C Cook - Responds to 05/14/1795 Letter Regarding Secondary System Fluid Flow Instability in Pwr'S (Characterized as Water Hammer)

ML18219D053
Person / Time
Site:	Cook
Issue date:	07/16/1975
From:	Hunter R American Electric Power Service Corp
To:	Kniel K Office of Nuclear Reactor Regulation
References
Download: ML18219D053 (14)

Text

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NRC DIS1 P)UTION FOB PART 50 DOCKET .. EBIAL (TEMiPOBARY FORM)

CONTROL NO 7571 FILE'ROM.

American electric Power 'ATE OF DOC DATE R EC'D LTR TWX OTHER Service Corp Broadway, N Y.

R.S ~ Hunter 7<<14<<75 TO: ORIG CC OTHER SENT t:RC PD

~ Mr Karl Kniel 1-signed SENT LOCAL PDR~~~~

CLASS UNC LASS PROP INFO INPUT NO CYS REC'D DOCKET NO:

XXXXX L 50-315 and 316 DESCR IPTION: ENCLOSURES:

Ltr re our 5-14-75 ltr ~,trans the following'.

~ ~ ~ Responses concerning secondary system fluid flow instability in PMR's ( characterized as "water hammer") .~ ,, ~ with attached j>CK o<<FDG~D dawings ~ ~ ~ ~ ~ ~ ~

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Docket, No. $0-315 end 50-31 DPN No. 58 a))d CPPR No. 61 ggpy~r Mr. Karl Kniel, Chief ping<(e~ 4i,ger g Light Mater Reactors /Pd Branch No. 2-2 io Division of Reactor Licensing U.S. Nuclear Regulatory Commission Nashington, D.C. 20555

Dear Mr. Kniel:

Mr. John

'n response Tillinghast to your letter of May 14, 1975 to concerned w'ith secondary system fluid flow instability in PNR's (characterized as "water hammer"),

we are attaching analyses and other relevant. information needed to determine the potential for occurrences and the potential consequences of such an event at the Donald C. Cook Nuclear Plant. These analyses are submitted by American Electric Power Service Corporation in direct response to the questions which were asked in the enclosure to your May 1+, 1975 letter. Please note that the Donald, C. Cook Nuclear Plant is designed, so that in the event water hammer effects are induced by uncovering of the feedwater sparger line, the piping stresses resulting from such an event will not exceed,.

the yield. stress for the feedwater piping.

Very truly yours, RSH:ma R. S. Hunter Attachment CC ~ John Tillinghast R. tj. Jurgensen - Bridgman G.

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'571 P. 11. Steketee R. J. Vollen Ro 't)lalsh

g.l: Describe all operating occurrences that could cause the level of the water/steam interface in the steam generator to drop below the feedwater sparger or inlet nozzles and allow steam to enter the sparger and/or the feedwater piping.

Answer: The items listed below are possible operating occurrences that could cause the level of the water/steam interface in the steam generators to drop below the feedwater sparger or inlet nozzle and allow steam to enter the sparger and/or the feedwater piping.

1. Main feedwater pumps trip on loss of or low suction pressure.

2. Main feedwater pump turbine trip on loss of or low feedwater pump turbine condenser vacuum.

3. Rapid surging of steam generator water/steam interface level caused by pressure transient in main steam system following reactor trip and subsequent main turbine trip.

4. Misoperation of either the auxiliary or main feedwater system controls during a plant startup or return to power when controls are in manual.

5. Main turbine trip followe'd by feedwater system isolation without either automatic auxiliary feedwater pumps starting or a reactor trip.

6. Failure of the feedwater controls without feedwater/main steam flow mismatch interlock causing a reactor trip.

g.2: Describe and show by isometric diagrams, the routing of the main and auxiliary feedwater piping from the steam generators outwards through containment, up to .the outer containment isolation valve and restraint. Note all valves and provide the elevations of the sparger and/or inlet nozzles and all piping runs needed to perform an independent analysis of drainage characteristics.

Answer: The attached Figures 1-5796-4 and MSK-718 provide the requested information for Donald C. Cook Nuclear Plant, Unit No. 1. Information for Unit No. 2 is not shown in that the only major difference between feedwater piping for both units is the orientation of piping. The basic dimensional configurations and piping elevations of concern with respect to "water hammer" analyses are identical in both units.

g.3: Describe any "water hammer" experiences that have occurred in the feedwater system and the means by which the problem was permanently corrected.

Allswer: There were no "water hammer" experiences that have occurred in the feedwater system as of the date of this response that required permanent correction.

There was one incident during the plant startup program when the feedwater sparger was uncovered due to the loss of offsite power test which was conducted.

There was no observable evidence of water hammer damage that resulted from this incident.

g.4'. Describe all analyses of the feedwater and auxiliary feedwater piping systems for which dynamic forcing- functions were assumed., Also, provide the results of any test programs that were carried out to verify that either uncovering of the feedwater lines could not occur at your facility, or if it did occur, that "water hammer" would not occur.

a. If forcing functions were'ssumed provide the technical bases that in analyses, were used to assure that an appropriate choice was made and that adequate conservatisms were included in the analytical model.

b. If a test program for assuring that was followed, provide the basis the program adequately tracked and predicted the flow instability event that occurred, and further, that the test results contained adequate conservatisms and an acceptable factor of safety, i.e., range of parameters covered all conceivable modes of operation.

c~ If neither your basis

a. nor b. has been performed, present for not requiring either and your plans to investigate this potential transient occurrence.

Answer: The attached memorandum entitled "Steam-Mater Slugging "in Steam Generator Feedwater Lines" by >lestinghouse Besearch Laboratories dated January 2, 1975 (Code No. 74-7E9-FLINE-Ml) provides the basis for the analysis of the feed-

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water system for fluid flow instability (water hammer.)

Specific analysis were not performed for the auxiliary line piping systems since water hammer due to the feedwater sparger uncovering could not be induced in those lines due to their location on the feedwater system.

A specific test program to determine the characteristics of flow instability due to the feedwater sparger uncovering was not performed on the Donald C. Cook Nuclear Plant.

The attached Westinghouse memorandum shows that water hammer effects are functions of the horizontal feedwater length from the internal tee center line to the point where the first downward turning elbow starts. The Donald C. Cook Nuclear Plant feedwater design for each of the four steam generators have zero horizontal length between the first downward turning, elbow and the steam generator nozzle. Therefore, from the standpoint of water hammer -effect due to uncovering of the sparger loop, the Cook Plant feedwater design is the optimum possible.

Analyses have been performed using the information contained in the attached Westinghouse memorandum to show that the secondary side fluid instability will not result in stresses which exceed yield.

Figure 9.4-1 presents a schematic representation of the feedwater piping on the Donald C. Cook Nuclear Plant showing the horizontal feedwater line length from the internal tee center line of the sparger loop to the point wh'ere the first elbow starts turning downward. This distance is approximately four feet. From information cited in the attached memorandum on Steam Water Slugging, it is concluded that the maximum water hammer forces induced in the piping would be as follows:

Maximum overpressure after bubble collapse:

1325 psi Peak Acoustic Overpressure:

940 psi Total Energy in Traveling Nave:

24,000 inch-pounds Duration of Pressure Nave 1.6 milliseconds Figure 9.4-1 shows that the horizontal length of feedwater piping of concern consists of 16" schedule 40 pipe material. The two pieces are joined by a butt weld using E7018 rod. Table 9.4-1 shows the yield stress of the various materials.

The peak hoop stresses, for the 1325 psi over-pressure was calculated for the 16" pipe, the 16" elbow and the welded joint. The results of these calculations are presented as the ratio of this peak stress to the yield stress in Table Q.4-1 also.

Peak hoop stresses in the steam generator nozzle and in the remainder of the internal feedwater ring are equal to, or smaller than stresses cited in Table Q.4-1.

A calculation was also performed conservatively assuming that all of the slug impact energy is contained within the 16" elbow. The results of the analysis show that this would result in peak hoop stresses of 21,880 psi, which is below the yield stress for the elbow material. A similar calculation assuming the slug impact energy is transmitted to the butt weld results in a peak hoop stress of 32,583 psi which is well. below the yield strength for this material.

/These analyses have shown that the average membrane /stresses during a postulated water hammer due to uncovering of the feedwater sparger are below the yield values in all cases. Thus it is concluded that the effects of water hammer would not cause the inadvertent rupture of a feedwater line at the Donald C. Cook Nuclear Plant.

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Donald Table Q.4-1 C. Cook Nuclear Plant Calculated Stresses Due to Feedwater t Hammer Piping Piece from Figure Q.4.-1) Material Cal. Max.

Stress (psi)

'ield (psi)

Stress Gale. Stress Yield Stress (g)

1) 16" L.R. Elbow-Sch 80 A106B 20,680 29)200 71

2) 16" Sch 40 pipe A106B 20 ~ 670 29~ 200 71

3) Butt 'tjlteld E7018 rod 30@96 58,ooo 53

4) Steam Generator Feedwater Nozzle A508CL2 Less than or 43,900 Less than equal to 1) 1) and 2) and 2) above above

g.5: Discuss the possibility of a sparger or nozzle uncovering and the consequent pressure wave effects that could occur in the piping following a design basis loss-of-coolant accident, assuming'concurrent turbine trip and loss of off-site power.

Answer: A design basis loss-of-coolant accident will cause initiation of a safety 4njection and containment isolation. As part of these sequences main feedwater isolation occurs. Depending on the size of the loss-of-coolant accident, the steam generator feedwater sparger may be uncovered; i.e.,

auxiliary feedwater if the system LOCA size is small, the will supply'ufficient water to prevent sparger uncovering. However, for most assumed break sizes, the sparger ring will be uncovered.

The consequent pressure wave effect of the uncovering of the sparger followed by bubble collapse is discussed in the Westinghouse Research Laboratories Research Memo 74.-7E9-FLINE-Ml entitled 'Steam-Water Slugging in the Steam Generator Feedwater Lines"which is enclosed. As is shown by the analysis presented previously, these pressure waves have no detrimential effect on the main feedwater system piping designed and installated at the Donald C. Cook Nuclear Plant.

9.6: If plant system design changes have been or are 'planned to be made to preclude the occurrence of flow instabilities, describe these changes or modifications, and discuss the reasons that made this alternative superior to other alternatives that might have been applied. Discuss the quality assurance program that was or will be followed to assure that the planned system modifications will have been correctly accomplished at the facility. If changes are indicated to be necessary for your plant, consider and discuss the effects of reduced auxiliary feedwater flow as a possible means of reducing the magnitude of induced pressure wave's, including positive means (e.g. interlocks) to assure sufficiently low flow rates while still meeting the minimum requirements for the system safety function.

Answer: No plant system changes are planned to preclude the occurrence of flow instabilities. As is shown in the responses to guestions l through 5, the feedwater system of the Donald C. Cook Nuclear Plant is designed in such a manner so as to assure any stresses created by flow instabilities due to feedwater sparger uncovering will not result in piping stresses above the yield stress.