ML20137H994

Forwards Response to Questions 12 & 13 of 850219 Request for Addl Info Re Pressurizer Safety & Relief valves.Loss-of-load Locked Rotor Event Produced Limiting Condition for Steam Discharge

ML20137H994
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	01/09/1986
From:	Leblond P COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 1076K, NUDOCS 8601220170
Download: ML20137H994 (23)

Text

--

N Commonwealth Edison

~.) One First N:tioritt Plaza. Chictgo. liknois O Addrzss Reply to. Post Offica Box 767

, Chicago, filinois 60690 January 9, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Conunission Washington, DC 20555

Subject:

Zion Nuclear Power Station Units 1 and 2 NUREG-0737, Item II.D.1 NRC Docket Nos. 50-295 and 50-304 References (a): February 19, 1985 letter from S. A. Varga to D. L. Farrar.

(b): June 18, 1985 letter from P. C. LeBlond to H. R. Denton.

Dear Mr. Denton:

Reference (a) contained 14 questions concerning Zion's pressurizer safety and relief valves. Reference (b) transmitted Commonwealth Edison Company's response to question numbers 1 through 11 and 14. The remaining responses to questions 12 and 13 are enclosed with this submittal. These responses were prepared for Commonwealth Edison by the engineering firm of O'Donnell and Associates, Inc. (ODAI).

If you have any further questions regarding this matter, please contact this office.

Very truly yours, P. C. LeBlond Nuclear Licensing Administrator ,

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cc NRC Resident Inspector - Zion J. A. Norris - NRR a0 ss 8601220170 e60109 PDR ADOCK 05000295 ppg 1016K p

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ODAI RESPONSE TO NRC QUESTIONS RELATED TO THE THERMAL l HYDRAULIC ANALYSIS OF THE INLET AND DISCHARGE PIPING MRC QUESTION 12 The submittal states that a thermal hydraulic analysis of the safety / relief l valve piping systes has been conducted, but does not present details of the ar.alysis. To allow for a complete evaluation of the methods used and the results obtained from the thermal hydraulic analysis, provide a discussion on the thermal hydraulic analysis that contain at least the following informations

a. Evidence that the analysis was performed on the fluid transient cases producing the maximum loading on the safety /PORY piping system. The cases should bound all steam, steam to water, and water flow transient conditions for the safety and PORY valves.

ODAI RESPONSE A generic analysis of the valve inlet fluid conditions for Westinghouse f plants is given in References 1 and 2. These studies clearly indicate that '

the most severe rate of pressurization and the highest pressure result from the locked rotor and loss-of-load events, respectively. The possibility of '

liquid being vented through the valves for a feedvater line break, for an [

extended high-pressure injection event, or for a cold over pressurization '

transient event are discussed on a generic basis for Westirghouse plants in References 1 and 2. A Zion plant-unique probabilistic assessment of the possibility of the flow of liquid through the valves was performed in Reference 3. This study determined that the total probability of liquid flow through the valves for these events is on the order of 1 x 10" . These  !

analyses considered single active falure and single operator error in the  !

determination of the event probability. The discharge of liquid from safety and relief valves in the Zion plant has been shown to be an entremely j unlikely event. The estimated frequencies are based upon conservative date <

end assumptions, .A they are sufficiently lov that even order-of-sagnitude  ;

errors would not .ffect the qualitative conclusion. '

The question of whether the pressurizer liquid level increases because of spray actuation (therefore representing some potential for safety and relief valve liquid discherge) was considered from a qualitative standpoint. The  ;

perspective is that it is extremely unlikely that any spray-induced level increase would be sufficient to actually result in liquid discharge through the safety or relief valves. The effect of the sprays is to condense steaa '

in the pressurizer and to thereby, in the majority of cases analyzed, curtail the overpressure transient before the safety or relief valve actuation preneure is recognized. For any resetning cases in which safety '

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and relief valve actuation cannot be entirely ruled out, the liquid level contribution of the pressurizer sprays is not expected to be severe enough to produce liquid discharge through the safety and relief valves. It may be desirable to confirm this assessment by a quantitative study; however, it is not anticipated that the renults of this study would reveal any spray-induced liquid discharge scenarios that are more likely to occur than those already analyzed in Reference 3.

The analyses of References 1, 2 and 3 serve as the basis of neglecting the transients that result in liquid flow through the valves. As a result of the above considerations. it was determined that the loss-of-load or locked rotor event produced the limiting conditions for steam discharge through the safety and relief valves.

NRC. QUESTION 12. PART b A detailed description of the methods used to perform this analysis. This includes a description of methods used to generate fluid pressures and moments over time and methods used to calculate resulting fluid forces on the system. Identify the computer progress used for the analysis and how these programs were verified.

1 ODAI RESPONSE The method used to perform tais analysis is based on the method used in the many thermal hydraulic analyses of safety / relief valve piping systems as found in the literature (see References 4, 5, 6, 7, 8 and 9). The steps used are given below

1. Develop an ANSYS (Reference 10) finite element structural model of the piping system. (See details below in ODAI response to NRC question 13). This is the most logical first step because the ultimate goal of the entire analysis is to verify that the stress levels in the piping system are in compliance with the ASME Boiler and Pressure Yessel Code (Reference 11). Therefore, the thermal hydraulic model which provides the input to the structural model must be compatible with the structural model. To this end, the guidelines given in References 12 and 13 were followed.

2. Develop a RELAPS/ MOD 1 (Reference 14) finite difference model of the piping system following the guidelines of References 12 and 13.

RELAPS/ MOD 1 vos written to investigate the thermal hydraulic reponse of light water reactors to a loss-of-coolant accident (LOCA). Its original intent van not for the determination of pressure waves in piping systems. As a result it has capabilities which are not necessary for the solution of pressure surge problems, e.g. RELAPS/ MOD 1 contains internal heat generation and reactor kinetics data which are not needed in relief valve applications. RELAPS/ MOD 1 does not give reaction forces due to

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pressure surges. Therefore a post-processor such so REPIPE (Reference 15) is needed to accept the RELAPS/N001 thermo hydraulic output and resolve it to forces. The application of RELAPS/N001 to problems similar to the relief valve discharge line probles (the safety / relief valve discharge in pressurized water reactore) has been discussed before (Reference 8). Control Data Corporation (CDC) esintaine RELAPS/ MOD 1 on its CYBERNET system and has written the post processor REPIPE which determines the fluid forces.

3. Using the thereal hydraulic output from RELAPS/N001 as input to REPIPE, obtain the force time history which is applied to the ANSYS structural model of Step 1. REPIPE le a post processor computer code that converto the thermal hydraulic output of RELAPS/N001 into force time histories at desired locations. These force time histories are generated between two arbitrary Junctions in the thermal hydraulic model. The conson analytical approach to to perfore structural evaluations with the forces acting along the axis of piping elemente. Therefore, the RELAPS/N001 thermal hydraulic output veo converted into forces using REPIPE. The force time histories generated by REPIPE vere composed.of the sus of the wave and blowdown forces. Reference 15 discusses the calculations of each of these forces. The REPIPE output consiste of the x, y and z componente of the sus of these forces relative to the absolute coordinate system used in the ANSYS structural model.

MRC QUEST 10M 12. PART c Identification of important parametere used in the thermal hydraulic analyste and rationale for their selection. These include peak pressure and pressurization rate, valve opening time, and fluid conditions at valve opening.

ODAI RESP 0ESE The generic analysis (References 1 and 2) for the four-loop plant predicted a peak pressure of 2SSS pela for the loss-of-load case. The Zion plant specific lose-of-load analyste presented in the FSAR determined the peak pressure to be 2332 peia. The manteve preneurization rate results from a locked rotor event. For the generic analysis, a pressurization rate of 144 pel/soe to predicted, while the Zion FSAR analysis estimates the pressurination rate to be 80 poi /sec. These analyses also confirmed that only stese le vented from the pressuriser in these cases.

The actual valve stes position (of the safety valve) wereue time for EPRI/CE Test 917 (Reference 2) le shown in Figure 1. The time history consiste of two distinct periode, the simmering time period and ' pop' time period. For Test 917, a staaering time of 0.9077 seconde and a pop time of 0.01475 seconde were sensured. The valve fully opened upon steam flow after the loop seal water had cleared the valve se a result of the elesering process.

4 l

l The valve opening characteristics employed in the RELAPS/N001 valve ocdel are superimposed over the data in Figure 1. The valve model used in the analysis employed conservative values of 0.88 second and 0.0145 second for j the siseer and pop periods. The fluid conditions in the RELAPS/N001 model

were based on the actual plant data as obtained from Reference 16. These conditions are given la Table 1 and Figure 2. The loop seal water was modeled with all the water in place upstrees of the valve.

MRC QUESTION 12. PART d An explanation of the method used to treet valve resistances in the analysis. Report the valve flow rates that correspond to the resistances used. Because the ASME Code requires derating of the safety valves to 90X of actual flow capacity, the safety valve analysis should be based on flows equal to 111X of the valve flow rating, unless another flow rate can be

. Justified. Provide information explaining how dorating of the safety valves

! was handled and describe methods used to establish flow rates for the safety i valves and PORVs in the analysis.

ODAI RESPQRSE I The valves were modeled using the conventional RELAP5/N001 valve component.

For this component a full open flow area of 0.025 ft2, a valve discharge  !

coefficient (C D of 0.8 and the opening time given in Ites 12c above were used. The results of the model gave a steady state stems flow rate of 129.3 lb,/sec which corresponds to 111X of the valve flow rating trating is 420,000 lb,/hr). The flow rate (e) at any instant of time is detereined by the following equations l e = AC DV2 PAP

l where A is the flow area, DC is the valve discharge coefficient, p is the density and AP is the preneure drop through the valve. By using the values l

of A, C D, p an AP calculated by REAPS /N001, one oMains the same value of a as calculated by RELAPS/M001.

MRC QUESTION 12. PART e l

A discussion of the sequence of opening of the safety valves that was used i to produce worst case loading conditions.

ODA! RESEQNSE

[ This has been covered in Items 12b and 12c above, i

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~9 MRC QUESTIDES 12. PART d A sketch of the thermal hydraulie model showing the size and number of fluid control valuees.

QDALRESPQRSE See Figures 2 and 3.

MRC QUE$H OS 12. PART a later. '

ERC QuRCTION 13 The submittal states that a structeral analysis of the safety /PORY valve

. piping system has been conducted. but does not present details of the unelysis. To allow for a complete evaluation of the methods used and results obtained from the structural unelysis. please provide reporte containing at least the icllowing informations

e. A detailed description,of the methods used to perfore the e analysis. Ide ntify the computer progress used for the analysis and her these progrees were varified.

QDAUlfSPGESE As sentioned in Ites 12b above, an ANSf5 finite element structural model was developed for the safety / relief valve piping systes. The ANSYS computer progree le a large-scale, general perpees computer progree for the solution of several cissees of engineering analysew. Analysis capabilities include static and dyneeles elasti; and plastics small and large deflootione; linear and non-linear. The estrix displacement method of analyses based upon finite element idealization is used. The library of finite eleopate includess elastia pipe, tee, elbow, bene and shell elementes pleetic pipe elbov, bene and shell elementes substructure (superelementals springlease viesents. The loading on the structure may be in the form of forces, displeuesente, pressures, temperatures or response spoetra. ANSYS has been verified and quality soeured for Nuclear Safety l Nelated analysee.

For this structitral analysis, the straight pipe sections were modeled as elastic pipe eleaente, the pipe teos wase modeled es elagtic pipe tee elemente, the valves and pipe supports were mode 14d as explained below in Ites 13c, the pipe elbove vese modeled as elastic pipe elbow elemente except that the three 12 inch diameter elbows in the header at the expected high streme locations were modeled se superelements which were obtained from a detailed elastic shell element model of the 12 inch diameter elbow.

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NRC QUESTION 13b l A description of the method used to apply the fluid forces to the structural model. Since the forces acting on a typical pipe segment are composed of a net, or " wave", force and opposing " blowdown" forces, describe the methods for handling both types of forces.

ODAI RESPONSE t

ANSYS has the capability to allow the structural analyst to apply the fluid i forces directly to the uodal locations of interest (i.e. locations of high stresse levels. ) As sentioned above in Ites 12b and Ites 13a, the guidelines given in Reference 12 and 13 were followed in developing the structural model so that the ANSYS model nodes included the locations of high stress

. levels. As mentioned in Ites 12b, REPIPE calculated the wave and blowdown forces for the desired locations and then the force time history was applied i to the ANSYS structural model in order to determine the stress levels of the l discharge piping system.

MRC QUESTION 13a A description of methods used to model supports, the pressurizer and relief j tank connections, and the safety valve bonnet assemblies and PORY actuator.

j ODAI RESP 0MSE Standard structural modeling practices were followed in developing the ANSYS i structural model of the discharge piping system. These include the j followings l 1. Pipe Supports The ease of the support clamps and the mass of the dynamic portion of the support attached to the pipe were modeled as a lumped mass i and placed on the pipe node at or very near to its physical 1

location. The values for the masses were obtained from References

! 17 and 18. A node at its physical location corresponding to the i centerline of the pipe was used to represent the end of the support i attached to the pipe. A node at its physical location was used to l represent the end of the support not attached to the pipe. This  ;

node was constrained in all degrees of freedom. An ANSYS spring l element was used to connect the two nodes of the support. The j values for the spring constants were obtained from References 19 i and 20.

i The constant force supports were modeled as a lumped mass to represent the pipe clamp and the dynamic portion of the support.

The values of the forces and the masses were obtained fros

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References 17 and 18. The masses and forces were placed on the pipe nodes at or very near to their physical locations.

2. Press.urizer and Relief Tank Connections j The locations of the pressurizer and relief tank connections were represented by pipe nodes at their physical locations corresponding to the centerline of the pipe. These nodes were constrained in all degrees of freedos.

3. Valves All of the valves were modeled using three relatively stiff beau elements and a mass element at the valve center of gravity as follows: one bene element running from the node at the valve inlet to the node at the valve outlet, one beam element running from the node at the valve inlet to the node at the valve center of gravity and one beau element running from the node at the valve outlet to the node at the-valve center of gravity. The values for the locations of the nodes at the center of gravities, inlets and l

outlets and the values for the masses were obtained from References 20 and 21.

4. Safety Valve Stands The safety valve stand was modeled as a two noda beau element. One node was attached to the center line of safety valve inlet piping corresponding to its physical location., The other node was attached to the center line of safety valve inlet piping corresponding to its physical location. The other node was at the enchor end of the stand at its physical location and was constrained in all degrees of freedom.

NRC QUESTION 13d An identification of the load combinatic'ns performed in the analysis together viih the allowable stress limits. Differentiate between load combinations used in the piping upstream and downstream of the valve.

Explain the mathematical methods used to perform the load combinations, and identify the governing codes and standards used to determine pipir.g and support adequacy.

ODAI RESPONSE The previous analyses reported in References 9 and 19 showed that an overstressed condition would result only in the event of the simultaneous opening of all three safety valves (see Ites 12 above for description of this event). Because this type of event is classified as an occasional load, Article F-1000 of the ASNE Boiler and Pressure Vessel Code (Reference

• 1 6

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22) applies. That is, the load combination consists of the sua of the sustained loads during normal plant operation (except peak pressure is used) and the dynamic load caused by the slug flow. Because the piping upstream of the safety valves is Seismic Class I piping, the dynamic force produced by a maximum credible earthquake (SSE) must also by included in the load combination. For this load combination the square root of the sua of the squares (SRSS) sethod is used. From F-1331.1 of Reference 22, the following standards were used to determine the piping and support adequacy:

1. The general primary membrane stres intensity P, shall not exceed the lesser of 2.4 S,and 0.7 S, for materials included in Table I-1.2, or 0.7 S, for materials included in Table I-1.1.

2. The local primary membrane stress intensity P shall g not exceed 150% of the limit for general primary membrane stress intensity P,.

3. The primary membrane (general or local) plus primary bending stress

intensity, Pg + P , shall be limited in accordance with one of the B

following provisions:

a. Stress intensity, Pg + PB, shall not exceed 150% of the limit for general primary membrane stress intensity P,;

b. static or equivalent static loads shall not exceed 90% of the limit analysis-collapse load using a yield stress which is the lesser of 2.35,and 0.7S , or 100% of the plastic analysis-collapse load or test collapse load (F-1321.6).

4. The average primary shear stress across a section loaded in pure l shear shall not exceed 0.42S .

For the materials under consideraiton in this analysis, values of the ultimate stress (S ), and yield stress (S,), and allowable stress (Sg ) as a function of temperature are given in Table 2.

t MRC QUESTION 13e An evaluation of the results of the strbctural analysis, including identification of over stressed locations and a description of modifications, if any.

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9 ODAI RESP 0MSE (Later)

NRC QUESTI0M 13f A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces.

ODAI RESPONSE See Figure 4.

NRC QUESTION 13a A copy of the contractors structural analysis report.

ODAI RESPONSE Later.

10 l

l TABLE 1 Fluid Initial Conditions Itests) Conditions

1. Fluid in pressurizer and fluid Saturated steam at the safety upstream of the loop seals valve set point pressure (2499.7 paia)

2. Fluid in the loop seals See Figure 2 for temperatures

3. Fluid downstream of the safety Air-water sixture at 100%

valves and inside the relative humidity at 110'F containment and 14.7 psia

4. Fluid outside of the containment Air-water mixture at 100%

but not in the relief tank relative humidity at BO'F and 14.7 psia

5. Fluid inside the relief tank Water at 80'F, air at 80'F

11 TABLE 2 Material Data Nominal Wall Size Pipe Thickness Temperature Location (in.) Schedule (in.) Neterial Ranae ('F)

Upstream of Safety / Relief Valves 6 160 0.718 SA-376 TP316 120-668 Upstream of Relief Valves 3 106 0.437 SA-376-TP316 120-668 Downstream of Relief Yalves 3 40 0.216 SA-312 TP304 110 Downstream of Safety Reliet Valves 6 40 0.280 SA-312 TP304 110 Header 12 40 0.406 SA-358 316 80-110 SA-312 TP304 SA-376 TP316 AND SA-358 316 Temp. S Sg S 8 A u M A

(*F) ikg11 (kei). (kei) (ksi) (ksi) (kei) 100 75.0 30.0 52.501 75.0 30.0 52.501 200 71.0 25.0 49.701 75.0 25.8 52.501 300 66.0 22.5 46.201 73.4 23.3 51.381 400 64.4 20.7 45.081 71.8 21.4 50.261 500 63.5 19.4 44.451 71.8 19.9 47.762 600 63.5 18.2 43.682 71.8 18.8 45.122 650 63.5 17.9 42.962 71.8 18.5 44.402 700 63.5 17.7 42.482 71.8 18.1 43.432 ISA = 0.7 S, 2S = 2.4 S A M 1

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. 19 REFERENCES

1. Westinghouse Nuclear Energy Systems, " Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program," WCAP-10105, June 1982.

2. Electric Power Research Institute, " Valve Inlet Fluid Conditions for Presurizer Safety and Relief Valves in Westinghouse-Designed Plants,"

EPRI NP-2296, EPRI Project V102-19, Final Report, December 1982.

3. Science Applications, Inc., "Probabilistic Evaluation of High Pressure Liquid Challenges to Safety / Relief Valves in the Zion, Byron /Braidwood PWR Plants," June 25, 1982.

4. House, R. K., et al., " Application of RELAPS/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads," EPRI NP-2479, EPRI Project V102-28, Final Report, December 1982.

5. Motloch, C. G., Van Blaricum, C. H., and Narum, R. E. ,

"RELAP5/ANSYR/ANSYS Hydrodynamic Force Calculation of the Electric Power Research Institute Safety and Relief Valve Discharge Test (CE Test No. 1027)," EI-83-12, December 1983.

6. Ca]igas, J. M., " Verification of the RELAP5-FORCE Hydraulic Force Calculation Code," Gilbert Associates, Inc., May 1984.

7. Semprucci, L. B. and Holbrook, B. P. , "The Appiciation of RELAP4/REPIPE to determine Force Time Histories on Relief Valve Discharge Piping,"

ASME, PVP-33, June 1977.

8. Strong, B. R., Jr. and Beschiere, R. J., " Steam Hammer Design Loads for Safety / Relief Valve Discharge Piping,' ASME, PVP-33, June 1977.

9. Sargent & Lundy Report SL 4283 dated May 2, 1984, " Evaluation of the Pressurizer Safety and Relief Valve Discharge Piping System - Zion Stations 1 and 2.*

10. ANSYS Engineering Analysis System, Revision 4.1, Swanson Analysis Systems, Inc., Houston, Pennsylvania.

11. " Power Piping," ANSI /ASME B31.1, ASME Code for Pressure Piping, The American Society of Mechanical Engineers, 345 East 47th Street, New York, New York, 10017.

12. Norton, P. J., " User's Manual for Program REPIPE," Utilities Service Center, CDC, Rockville, Maryland.

-l 20 t

13. " Criteria and Guidelines for the Design of Safety and Relief Valve i Installation in Westinghouse Pressurized Water Reactor Plants,'

Westinghouse Electric Corporation, NES, PWR Systems Division, October 1972.

14. Ransos, V. H., et al., 'RELAP5/M001 Code Manual," Volumes 1-2, NUREG/CR-1826, EGG-2070 Draft, Rev. 2, September 1981.

15. Norton, P. J., " User's Manual for Program REPIPE,' Utilities Service Center, CDC, Rockville, Maryland.

16. Graesser, K. L., (Zion Station Superintendent) to Butterfield, L. D.,

(CECO), Letter November 9, 1982, " Unit 2 Pressurizer Safety Valve Loop Seal Temperatures.*

17. Sargent & Lundy Reactor Coolant System Support Drawings:

1 I

Hanner No. Egtt Hanner No. Enig.

IRC146-FR1 8-25-77 RCH-1008 12-18-72 1RC146-SR1 4-21-77 RCH-1009 1-28-74 1RC147-SR1 4-21-77 RCH-1014 10-27-72

, 1RC147-SR2 4-21-77 RCRS-1112 11-20-72 l 1RC151-RV1 4-21-77 RCRS-1114 6-02-71 1RC157-RV1 8-25-77 RCRS-1115 11-20-72 1RC157-RV2 4-21-77 RCRS-1119 2-16-73 RCH-1005 10-27-72 RCRV-001 12-21-72 RCH-1007 1-12-73

, 18. Stone & Webster Bulletin 79-14 Modification Support Drawings:

llanner No. Dals.

RCH1006 2-10-81 RCH1010 1-30-81 RCRS1117 1-30-81 RCRS1118 1-30-81 I

RCRS1120 2-04-81 RCRS1121 1-30-81 RCRS1122 1-30-81 RCRS1123 2-04-81 RCRS1117A 7-22-81 RCRS1117B 7-22-81 RCRS1118A 7-22-81 RCS1011 --

RCS1012 --

RCS1013 --

4 r -- - . - , - . - - - - , - - - , - , - , , , , . - - .. . . n.-,,------,,-,----,-.----w- - - - - - - - - - -

r-

, 21 i

l

19. Books 1 through 6, inclusive, of Stone & Webster, " Zion Station Pipe  ;

Stress and Support Analysis Report," Number 13430RC - 2, 3, 4, 5, Revision  !

O, dated January 17, 1983, Cossonwealth Edison Job Order 13430.01 for Reactor Coolant (Pressurizer 1RC002 to Pressurizer Relief Tank 1RC003).

20. Sargent & Lundy Report No. 037064, Project No. 6320-00. "Dynesic Analysis of Typical Pressurizer Safety and Relief Valve Discharge Piping Due to Valve Actuation," dated August 1982.

21. Sargent & Lundy Drawing M-418, Pressurizer Piping Analytical Data Isometric, Zion Station Unit 1, Sheet No. 1, Rev. D, Dated July 31, 1979.

20. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendix F, 1983.