ML19262A270

General Info on Effects of Fluid Flow Instability in Main & Emergency Feedwater Systems of TMI-1, Submitted to Met Ed

ML19262A270
Person / Time
Site:	Three Mile Island
Issue date:	09/30/1975
From:	Rochino A, Snow R GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT
To:
Shared Package
ML19262A263	List: ML19262A262 ML19262A270
References
1881, NUDOCS 7910260602
Download: ML19262A270 (25)

Text

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THREE MILE ISLAND NUCLEAR STATION UNIT 1 GINIRAI DNFORMATDON ON ':'H3 EFFICTS OF F: UDD FLOW HNSTABDLOTY DN THI MADN AND IMISGENCY FE!DWATIP, SYSTIMS OF TH EI! MO:.2 05:.AN D NUCL3AR STAT:ON UND71 METROPOLITAN EDISON COMPANY SUBSIDI ARY OF GENERAL PUBLIC UTILITIES CORPORATION 1484 276 Gilbert Associates, Inc.

%,,, engineers andconsultants Reading. Pennsylvania

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SEPTEMBER, 1975 GAI REPORT NO. 1881 GENERAL IhTOPMATION ON THE EFFECTS OF FLUID FLOW INSTABILITY IN THE MAIN AhT EMERGENCY FEEDb'ATER SYSTEMS OF THREE MILE ISLAND NUCLFAR STATION UNIT 1 METROPOLITAN EDISON COMPANY Subsidiary of General Public Utilities Corporation i

Prepared by: h, w R.A. Snow, PE, Piping Engineer, Piping Engineering Dept.

Revie;ted and .

Approved by: '/ _ A A.P. Rocitin& PE, PhD, Supervisory Engineer, Piping Engineering Dept.

Gilbert Associa:es, Inc.

525 Lancaster Avende JkT Reading, Pennsylvania CILDERT 4550C3 %TES. INC.

TABLE OF CONTENTS Section Title _Page

1.0 INTRODUCTION

1 2.0 NUCLEAR REGULATORY COMM'_SSION GUESTIONS AND CONSEQUENT RESPONSES 2 3.0 MAIN AND DERGENCY FEEDWATER TO STEAM GENERATOR 1A 8 3.1 MAIN FEEDWATER SYSTEM 9 3.2 EMERGENCY FEED' WATER SYSTEM 10 4.0 MAIN AND EMERGENCY FEEDWATER TO STEAM GEhERATOR 1B 12 4.1 MAIN FEEDWATER SYSTEM 12 4.2 EMERGENCY FEEDWATER SYSTEM 14

5.0 REFERENCES

16 Fr uRES FRure No. Title 3-1 MAIN AND DERGENCY FEEDWATER ELEVATIONS FROM STEAM GENERATOR 1A TO ANCHORS PAST CONTAIIOENT ISOLATION VALVES 3-2 MAIN AND EMERr.ENCY FEEDWATER DF~ AILS FROM STEAM GENERATOR 1A TO REACTOR BUILDING PENETRATIONS NO. 227 AND NO. 110 3-3 MAIN AND DERGENCY FEEDWATER DETAILS FROM REACTOR BUILDING PENETRATIONS NO. 227 AND NO. 110 TO ANCHORS PAST CONTAINMENT ISOLATION VALVES 4-1 MAIN AND DERGENCY FEEDWATER ELEVATIONS FROM STEAM GENERATOR 1B TO ANCHORS PAST CONTAI! DENT ISOLATION VALVES 4-2 MAIN AND EMERGENCY FEEDWATER DETAILS IROM STEAM GENERATOR 1B TO REACTOR BUILDING PEFETRATIONS NO. 103 AND NO. 111 4-3 MAIN AND DERGENCY FEEDWATER DEIAILS FROM REACTOR BUILDING PENETRATIONS NO. 103 AND NO. 111 TO ANCHORS PAST CONTAI:OENT ISOLATION VALVES 1484 278 CIL!lEltT 4HOCI LTES, I.%C.

1

1.0 INTRODUCTION

This report presents the results of a design review and evaluation performed in response to the United States Nuclear Regulatory Co= mission (NRC) request, " General Information Required for Consideration of the Effects of Secondary System Fluid Flow Instability," dated Fby 15, 1975, for Three Mile Island Nuclear Station, Unit 1 (TMI-1).

Included are complete responses to specific t:RC questions (Reference 1) considering design and reference data requested by the NRC. Most of these respcases are based heavily upon information provided in Reference 2.

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2.0 NC LEAR REGULATORY COMMISSION C:UESTIONS AND CONSECUENT RESPONSES Ouestion No. 1 Describe all operating cccurrences that could cause the level of the water / steam interface in the steam gtnerator to drop below the feedwater sparger or inlet nozzles, and allow steam to enter the sparger and/or feedwater piping.

Reseense The design of the Babcock and Wilcox (3&W) once through steam generator (OTSG) requires that the level of the water / steam interface remain below the feedwater inlet nozzles during operation.

However, the arrangement of the feedwater nozzles, external ring distribution header and feedwater piping leading up to the header is such that r. cam cannot enter the feedwater piping. The piping immediately outside the steam generator contains a " gooseneck" or trap arrangement which is always filled with watcr. This precludes any steam from entering the feedwater piping.

At TMI-l there is a small one-inch bypass line around the feedwater control valves which maintains a continuous flow rate to the steam generators prior to power operation. This small flow rate (which begins when the unit is cold) will, in itself, keep the feedwater lines leading up to the steam generator f 211 of water and preclude steam from entering the pipe. Once perar operation begins, the normal feedwater flow rate fills the pipes, feedwater distribution header and nozzles with water and they remain filled throughout power operation.

1484 2h80 cit.itEHT 650Ci tTES, ISC.

3 Question No. 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 nn independent analysis of drainage characteristics.

Response

The routing of the main and emergency feedwater piping from the steam generators to the anchors past the outer containment isolation valves is discussed in Sections 3.0 and 4.0. The isometric diagrams referenced in the discussions contain information relative to valve locations, elevations of spargers or inlet nozzles and the elevaticns of the different parts of the piping runs.

It should be noted that the main and emergency feedwater systems are completely separated.

Question No. 3 Describe any " water hammer" experiences that have occurred in the feedwater system and the means by which the problem was permanently corrected.

Response

No " water hammer" experiences have occurred in the main or emergency feedwater systems at TYI-l (Reference 4) .

1484 28i Cit.IIERT Addoll s TEd. INC.

4 Question No. 4 Describe all analyses of the feedwater and auxiliary 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 assumed in your analysis, provide the technical bases that 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 was followed, provide the basis for assuring that the program adequately tracked and predicted the flow instability event that occurred, and further, that the test reaults contained adequate conservatisms and an acceptable factor of safety, e.g., range of parameters covered all conceivable modes of operation.

c. If neither a nor b have been performed, present your basis for not requiring either and your plans to investigate this potential transient occurrence.

Response

Items a and b, above, have not been performed nor are future investigations concerning the potential occurrence of this transient planned for the following reasons:

a. No " water ha=mer" phenomena have occurred in TMI-1 main or emergency feedwater lines from the steam generator inlet nozzles or spargers to the pipe anchors past the outer containment isolation valves.

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b. Similar fluid flow transients have not been observed during

. operations at Oconee Unit 1, a nuclear unit similar to TMI-1.

c. System design and complete separation of main and emergency feedwater piping preclude the possibility of steam entering the feedwater lines. Consequently, the flow instability that generates the " water hammer" transient will not occur.

d. A one-inch bypass line is provided around each feedwater control valve which :'.ntains a continuous flow rate to the steam generators at all pcwer levels. This ensures that the feedwater piping is filled at all times.

Question No. 5 Discuss the possibility of a sparger or nozzle uncovering and the consequent pressure wave eff sets that could occur in the piping following a design basis loss-of-coolant accident, assuming concurrent trip and loss of offsite power.

Resoonse The steam generator water level is below the feedwater inlet nozzles during power opera; ion. However, steam will not enter the feedwater piping and no pressure wave effects will occur in the piping following a design basis accident with concurrent turbine trip and loss of offsite power due to the " gooseneck" arrangement of the feedwater piping directly outside the steam generator. Even when the main feedwater pumps trip and feedwater flow rate is zero, the trap remains full of water and precludes the possibility of any steam entering the piping.

1484 283 GILilEltT i.* p oCl \ TE.* 15 0.

6 A test was run at Oconee Unit 1 (a unit similar to TMI-1) from 40 percent power during which the main feedwater pumps and the turbine generator were tripped and auxiliary feedwater flow was initiated. The auxiliary feedwater system on the E&W OTSG (including distribution header and piping) is completely separated from the main feedwater system. The auxiliary feedwater enters the unit through a separate header at the top of the tube section. This test very closely simulates the effects of a loss of offsite power on the secondary plant. The steam generator and feedwater piping directly adjacent to the steam generator were monitored for noises using the B&W loose parts monitoring system. No unusual noises were heard, confirming the fact that no " water hammer" in the feedwater piping occurred during the test.

Question No. 6 If plant system design changes have been cr are planned to be made to preclude the occurrence of flow instabilities, describe these changes or modification ~s, 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 waves, including positive means (e.g., interlocks) to assure sufficiently 1484 284 GII.11E!!T 68001 \TE5. INC.

7 low flow rates and still meet the minimum requirements for the system safety function.

Resconse It has been shown in the preceding responses that system design already precludes the possibility of the occurrence of " water han=er". Therefore, no system design changes are planned for further protection against this occurrence. Furthermore, TMI-l is provided with a one-inch bypass line around the feedwater control valves which maintains a continuous low flow rate to the steam generators at all power levels. This feature, in conjunction with other inherent system designs, precludes the possibility of the occurrence of the " water hanmer" phenomenon.

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8 3.0 MAIN A','D EMERGENCY FEEDWATER TO STEAM CENERATOR 1A Figure 3-1 is an isometric representat?on of those portions of the main feedwater (FW) and emergency feedwa: er (EFW) systems associated wica steam generator LA. These systems are completely separated from each other. All valves are noted and the height above (+) or below (-) the datum lines for the different parts of the piping runs are indicated. The datum lines are the elevations of the injection nozzles which are as follows:

a. For FW 324 ft-7-ll/16 in

b. For EFW 337 f t-5-3/8 in Since the two systems are independent of each other, the indicated elevations for parts of the piping runs of each system are referenced to the elevations of the associated steam generator injection nozzles.

The information provided by Figure 3-1 is sufficient to perform an independent analysis of the drainage characteristics of the entire feedwater system associated with steam generator lA. This isometric diagram depicts the piping runs for each system (FW and EFW) from the steam generator inlet nozzles to the pipe anchor past the containment isolation valves outside the reactor building.

Figures 3-2 and 3-3 show the A tails inside and outside the reactor building, respectively, of the FW and EFW system piping. In the following discussion the height, or elevation, of each part of the piping run with respect to the datum lines will be indicated within parentheses using the sign convention noted, i.e., (+) for above and (-) for below the datum line elevations.

1484 286 cn.noiT swoa sTu. t>r. ,

9 3.1 MAIN FEEDWATER SYSTEM The FW system piping enters steam generator LA through 32 injection nozzles at elevation 324 ft-7-ll/16 in. Each of the 32 injection nozzles is connected by a 3 ft-9 in long, vertical section of 3 in pipe down to a 14 in diameter horizontal ring header (-3 ft-9 in).

The ring header consists of two half rings as shown by Figure 3-2.

Each half ring connects to a 14 in supply pipe which descends 4 ft-9 in (-8 ft-6 in) from the centerline of the ring. The two sections of 14 in pipe run horizontally for 19 ft-4-1/2 in and 20 ft-6 in, respectively, and then connect through 20x1,4 in reducers to a common 20x20x20 in tee. The 20 in diameter feedwater supply pipe then rises 40 f t-10-5/16 in (+32 f t-4-5/16 in), runs horizontally for 29 ft-1 in (+32 ft-4-5/16 in), descends 28 ft-0 in

(+4 f t-4-5/16 in) and runs horizontally for 17 f t-9 in (+4 f t-4-5/16 in) as shown by Figure 3-2.

The FW line is anchored to the reactor building at penetration no. 227.

In summary, there are 145 feet of FW piping inside the reactor building and there are no valves in the line. There are, however, 8 rupture restraints, 3 hydraulic snubbers and 7 spring and constant support hangers.

Figure 3-3 shows the FW piping run outside of the reactor building.

The 20 in pipe runs horizontally from penetration no. 227 for 16 f t-10 in (+4 f t-4-5/16 in) , descends 5 f t-0 in (-0 f t-7-ll/16 in) and runs horizontally for 2 f t-6 in (-0 f t-7-ll/16 in) before passing through an isolation check valve, FWV-12A.

1484 287 GII.0E!!T e500 STE5 INC.

10 The 20 in pipe continues to run horizontally for 12 ft-4 in

(-0 ft-7-ll/16 in) to the electric motor operated (EMO) 20 in feedwater control valve, FWV-17A. The FW line then rises 22 f t-0 in

(+21 ft-4-5/16 in). The main feedwater isolation valve, FWV-5A, is located in this vertical piping run. A 6 in bypass line around the feedwater control valve, FWV-17A, and the main feedwater isolation valve, FWV-5A, is installed to facilitate control of low and intermediate FW flow. This bypass line is provided with two control valves, FWV-92A and FWV-16A.

The feedwater control valve, FWV-17A, is provided with a 1 in bypass line to ensure continuous feedwater ficw during all modes of FW system operation.

The remainder of the 20 in FW line runs horizontally for 105 f t-10 in

(+21 f t-4-5/16 in) before descending 21 f t-9 in (-0 f t-4-11/16 in) and then runs horizontally for 4 ft-9-1/2 in (-0 ft-4-ll/16 in) to g, a pipe anchor.

3.2 EMERGENCY FEEDWATER SYSTEM As shown in Figure 3-2, the flow from the EFW system enters steam generator LA through seven injection nozzles at an elevation of 337 f t-5-3/8 in (the EFW datum elevation),12 f t-9-ll/16 in above the FW injection nozzles. Each of the seven injection nozzles connects to a 3 in pipe which descends 1 f t-8 in and joins a 6 in diameter, horizontal, 300 degree ring header (-1 ft-8 in). This ring header is supplied by a single 6 in pipe which descends 5 ft-9-3/8 in

(-7 ft-5-3/S in) from the ring centerline. This 6 in pipe then runs horizontally for 6 ft-3 in (-7 ft-5-3/8 in), rises ft-0 r I

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(-l f t-5-3/ 8 in) , runs horizontally for 34 f t-6 in (-1 f t-5-3/8 in) ,

descends 4 f t-0 in (-5 f t-5-3/8 in), runs horizontally for 60 f t-2 in

(-5 f t-5-3/8 in) , descends 14 f t-0 in (-19 f t-5-3/8 in) , runs horizontally for 18 f t-0 in (-19 f t-5-3/8 in) and is anchored to the reactor building at penetration no. 110.

In summary, there is 150 feet of EFW piping inside the reactor building and there are no valves in the line. There are, however, 2 guides, 8 hydraulic snubbers, 5 sway struts, 8 rupture restraints and 9 spring hangers and constant load supports.

Outside the reactor building, as shown by Figure 3-3, the 6 in EFW pipe runs horizontally from penetrati,n no. 110 for 13 ft-10 in

(-19 f t-5-3/8 in) passing through check va ve EFV-12A. The pipe then descends 10 ft-0 in (-29 ft-5-3/8 in), runs horizontally for 34 f t-11 in (-29 f t-5-3/8 in), descends 9 f t-8-9/16 in (-39 f t-1-15/16 in) and runs horizontally to the EFW control valve, EFV-30A. After the control valve, the pipe descends 2 f t-3-7/16 in and runs horizontally 8 ft-0 in to a pipe anchor.

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12 4.0 MAIN AND CIERGENCY FEEDWATER TO STEAM GENERATOR 1B Figure 4-1 is an isemetric representation of those portions of the main feedwater (FW) and emergency feedwater (EFW) systems associatec with steam generator 1B. These systems are completely separated from each other. Presentation of the heights of various parts of the piping runs follows the conventions discussed in Section 3.0.

The datum line elevations (injection nozzle elevations) are as follows:

a. For FW ,324 ft-7-11/16 in

b. For EFW 337 ft-5-3/8 in Figures 4-2 and 4-3 show the details, inside and outside the reactor building, respectively, of the FW and EFW piping.

4.1 MAIN FEEDWATER SYSTEM The FW system piping enters steam generator 1B through 32 injection

,_ nczzles at elevation 324 ft-7-ll/16 in. Each of the 32 inj ection nozzles is connected by a 3 ft-9 in long, vertical section of 3 in pipe down to a 14 in diameter horizontal ring header (-3 ft-9 in).

_ The ring header consists of two half rings, each of which is connected to a 14 in supply pipe which descends 4 ft-9 in

(-8 f t-6 in) from the centerline of the ring. The two sections of 14 in pipe run horizontally for 14 f t-il in and 17 f t-il in, respectively, and then connect through 20x14 in reducers to a

_ common 20x20x20 in tee. The 20 in diameter feedwater supply pipe then rises 14 ft-10-5/16 in (+6 ft-4-5/16 in), runs horizontally for 9 f t-9 in (+6 f t-4-5/16 in), descends 14 f t-10-5/16 in

(-8 f t-6 in), runs horizontally for 58 f t-4 'n (-8 f t-6 in),

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13 descends 13 f t-5-11/16 in (-21 f t-ll-11/16 in), runs horizontally for 46 f t-5 in (-21 f t-ll-ll/16 in), rises 13 f t-5-ll/16 in

(-8 f t-6 in) and runs horizontally for 5 f t-8-3/4 in to an anchor at reactor building penetration no. 103 as shown by Figure 4-2.

In summary, there are 186 feet of FW piping inside the reactor building and there are no valves in the line. There are, however, 9 rupture restraints, 5 hydraulic snubbers and 13 spring and constant support hangers.

Figure 4-3 shows the FW piping outside of the reactor building.

The 20 in pipe runs horizontally from penetration no. 103 for 21 f t-9-11/16 in (-8 f t-6 in) and rises vertically 7 f t-10-5/16 in

(-0 f t-7-ll/16 in) to an isolation check valve FWV-12B.

The 20 in pipe then runs horizontally for 10 ft-7-3/8 in

(-0 f t-7-ll/16 in) to the electric motor operated (EMO) 20 in feedwater control valve, FWV-17B. The FW line then rises 22 ft-0 in

(+21 ft-4-5/16 in). The main feedwater isolation valve, FWV-5B is located in this " cal pip ng run. A 6 in bypass line around the feedwater control ,alves, FWV-17B, and the main feedwater isolation valve, FWV-5B, is installed to facilitate control of low and ~

intermediate FW flow. This bypass line is provided with two control valves, FWV-92B and FWV-16B.

The feedwater control valve, FWV-17B, is provided with a 1 in bypass line to ensure continuous feedwater flow during all modes of FW system operation.

1484 291 Gil.DERT %$50C STE5. INC.

14 The remainder of the 20 in FW line runs horizontally for 52 ft-9 in

(+21 f t-4-5/16 in), descends 21 f t-9 ia (-0 f t-4-ll/16 in) and runs horizontally for 7 f t-9-1/2 in (-0 f t-4-ll/16 in) to an anchor.

4.2 EMERGENCY FEEDWATER SYSTEM As shown in Figure 4-2, the flow from the EFW system enters steam generator 13 through seven injection nozzles at an elevation of 337 ft-5-3/8 in (the EFW datum elevation), 12 ft-9-11/16 in above the FW injection nozzles. Each of the seven injection r.nzzles connects to a 3 in pipe which descends 1 f t-8 in and joins a 6 in diameter, horizontal, 300 degree ring header (-1 f t-8 in) . This ring header is supplied by a single 6 in pipe which descends 5 f t-9-3/8 in (-7 f t-5-3/8 in), runs horizontally for 6 f t-3 ia

(-7 f t-5-3/8 in), rises 6 f t-0 in (-l f t-5-3/8 in), runs horizontally for 11 f t-9-7/16 in (-l f t-5-3/8 in), descends 4 f t-3 in

(-5 f t-8-3/8 in), runs horizontally for 92 f t-3 in (-5 f t-8-3/8 in) ,

descends 13 f t-9 in (-19 f t-5-3/8 in) , runs horizontally for 40 f t-6 in (-19 f t-5-3/8 in) and is anchored to the reactor building at penetration no. 111.

In summary, there is 156 feet of EFW piping inside the reactor building and there are no valves in the line. There are, however, 3 guides, 2 hydraulic snubbers, 9 sway struts, 9 rupture restraints and 11 spring, constant load and rigid hangers.

Outside the reactor building, as shown by Figure 4-3, the 6 in EF'W pipe runs horizontally fro = penetration no. 111 for 14 ft-0 in

(-19 f t-5-3/8 in) passing through check valve, EFV-123. The pipe then descends 19 f t-8-9/16 in (-39 f t-1-15/16 in), runs horizontally cn.nEnT moci ms, t.sc. t4g4 }O}

15 for 9 f t-3 in (-39 f t-1-15/16 in) to the EFW control valve, EFV-30B, descends 2 f t-3-9/16 in (-41 f t-5-1/2 in) and runs horizontally for 2 f t-3 in (-41 f t-5-1/2 in) to a pipe anchor.

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5.0 REFERENCES

1. United States Nuclear Regulatory Commission letter and attachment to Metropolitan Edison Ccepany, Docket No. 50-289; requesting general information; Mayl5, 1975.

2. Babcock and Wilcox, " Prevention of Waterhammer", presentation meeting with the United States Nuclear Regulatory Comn.ission; April 1, 1975.

3. Babcock and Wilcox letter to Metropolitan Edison Company from Mr. J. D. Phinney to Mr. R. M. Klingaman; June 30, 1975.

4. Telecon memo, Mr. A. P. Rochino of Gilbert Associates, Inc.,

to Mr. G. Kunder of Three Mile Island Nuclear Station Unit 1, Operations; August 19, 1975.

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