ML20140E603
| ML20140E603 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 01/22/1986 |
| 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 8602040033 | |
| Download: ML20140E603 (16) | |
Text
n N
Commonwealth Edison
.,b One First futional Plaza, Chicago. Illinois C'
Address R: ply to: Post Office Box 767 g
Chicago, Illinois 60690 N
January 22, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 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.
(b): January 9, 1986 letter from P. C. LeBlond to H. R. Denton.
Dear Mr. Denton:
Reference (a) contained 14 questions concerning Zion's pressurizer safety and re.'ief 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 were enclosed with reference (c).
This submittal transmits revised responses to questions 12 and 13.
The allowable stress values contained in Table 2 have been corrected. Thus, please replace the responses transmitted with reference (c) with those enclosed with this letter.
If you have any further questions regarding this matter, please contact this office.
Very truly yours, P. C. LeBlond Nuclear Licensing Administrator lm ec: NRC Resident Inspector - Zion
(
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1 ODAI RESPONSE TO NRC QUESTIONS RELATED TO THE THERMAL HYDRAULIC ANALYSIS OF THE INLET AND DISCHARGE PIPING NRC OUESTION 12 The submittal states that a thermal hydraulic analysis of the safety / relief valve piping system has been conducted, but does not present details of the analysis. 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 information:
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 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 feedwater line break, for an extended high-pressure injection event, or for a cold over-pressurization transient event are discussed on a generic basis for Westinghouse 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 plar.t has been shown to be an extremely unlikely event. The estimated frequencies are based upon conservative data and assumptions, and they are sufficiently low that even order-of-magnitude errors would not affect 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 discharge) 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 ef fect of the sprays is to condense steam in the pressurizer and to thereby, in the majority 'of cases analyzed, curtail the overpressure transient before the safety or relief valve i
2 actuation pressure is recognized. For any remaining cases in which safety 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 results 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 descriptio~n of methods osed to generate fluid pressures and momenta over time and methods used to calculate resulting fluid forces on the system.
Identify the computer programs used for the analysis and how these programs were verified.
1 ODAI RESPONSE The method used to perform this 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 l
and Pressure Vessel 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 RELAP5/M001 (Reference 14) finite difference model of the piping system following the guidelines of References 12 and 13.
l RELAP5/M001 was written to investigate the thermal hydraulic l
resonse of light water reactors to a loss-of-coolant accident I
(LOCA).
Its original intert was 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. RELAP5/M001 contains internal heat generation and l
l
3 reactor kinetics data which are not needed in relief valve applications. RELAP5/M001 does not give reaction forces due to pressure surges. Therefore a post-processor such as REPIPE (Reference 15) is needed to accept the RELAP5/M001 thermo hydraulic output and resolve it to forces. The application of RELAPS/ MOD 1 to problems similar to the relief valve discharge line problem (the safety / relief valve discharge in pressurized water reactors) has been discussed before (Reference 8). Control Data Corporation (CDC) maintains RELAP5/ MOD 1 on its CYBERNET system and has written the post-processor REPIPE which determines the fluid forces.
3.
Using the thermal hydraulic output from RELAPS/M001 as input to REPIPE, obtain the force time history which is applied to the ANSYS structural model of Step 1.
REPIPE is a post-processor computer code that converts the thermal hydraulic output of RELAP5/ MODI into force time histories at desired locations. These force time histories are generated between two arbitrary junctions in the thermal hydraulic model. The common analytical approach is to perform structural evaluations with the forces acting along the axis of piping elements. Therefore, the RELAP5/M001 thermal hydraulic output was converted into forces using REPIPE. The force time histories generated by REPIPE were composed of the sum of the wave and blowdown forces. Reference 15 discusses the calculations of each of these forces. The REPIPE output consists of the x, y and z components of the sum of these forces relative to the absolute coordinate system used in the ANSYS structural model.
HRC QUESTION 12, PART c Identification of important parameters used in the thermal hydraulic analysis and rationale for their selection. These include peak pressure and pressurization rate, valve opening time, and fluid conditions at valve opening.
ODAI RESPONSE l
The generic analysis (References 1 and 2) for the four-loop plant predicted a peak pressure of 2555 psia for the loss-of-load case. The Zion plant l
specific loss-of-load analysis presented in the FSAR determined the peak pressure to be 2532 psia. The maximum pressurization rate results from a locked rotor event. For the generic analysis, a pressurization rate of 144 psi /sec is predicted, while the Zion FSAR analysis estimates the pressurization rate to be 80 psi /sec. These analyses also confirmed that only steam is vented from the pressurizer in these cases.
The actual valve stem position (of the safety valve) versus time for EPRI/CE Test 917 (Reference 2) is shown in Figure 1.
The time history consists of j
two distinct periods, the simmering time period and " pop" time period. For l
4 Test 917, a sinnering time of 0.9077 seconds and a pop time of 0.01475 seconds were measured. The valve fully opened upon steam flow af ter the loop seal water had cleared the valve as a result of the simmering process.
The valve opening characteristics employed in the RELAP5/M001 valve model 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 the simmer and pop periods. The fluid conditions in the RELAPS/M001 model were based on the actual plant data as obtained from Reference 16. These conditions are given in Table 1 and Figure 2.
The loop seal water was modeled with all the water in place upstream of the valve.
NRC QUESTION 12, PART d An explanation of the method used to treat 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 90%
of actual flow capacity, the safety valve analysis should be based on flows equal to 111% of the valve flow rating, unless another flow rate can be justified. Provide information explaining how derating of the safety valves was handled and describe methods used to establish flow rates for the safety valves and PORVs in the analysis.
ODAI RESPONSE The valves were modeled using the conventional RELAP5/M001 valve component.
For this component a full open flow area of 0.025 ft, a valve discharge coefficient (C )
f 0.8 and the opening time given in Item 12c above were D
used. The results of the model gave a steady state stems flow rate of 129.3 lb /sec which corresponds to 111% of the valve flow rating (rating is 420,000 lb /hr). The flow rate (m) at any instant of time is determined by the to 10 wing equation:
m = AC PO D
where A is the flow area, C is the valve discharge coef ficient, p is the D
density and AP is the pressure drop through the valve. By using the values of A, C, p and AP calculated by RELAP5/M001, one obtains the same value of D
m as calculated by RELAP5/ MOD 1.
NRC QUESTION 12, PART e A discussion of the sequence of opening of the safety valves that was used to produce worst case loading conditions.
5 ODAI RESPONSE This has been covered in Items 12b and 12c above.
NRC QUESTIONS 12, PART d A sketch of the thermal hydraulic model showing the size and number of fluid control volumes.
00AI RESPONSE See Figures 2 and 3.
NRC OUESTION 12, PART g Later.
NRC QUESTION 13 The submittal states that a structural analysis of the safety /PORV valve piping system has been conducted, but does not present details of the analysis. To allow for a complete evaluation of the methods used and results obtained from the structural analysis, please provide reports containing at least the following information:
a.
A detailed description of the methods used to perform the analysis.
Identify the computer programs used for the analysis and how these programs were verified.
ODAI RESPONSE As mentioned in Item 12b above, an ANSYS finite element structural model was developed for the safety / relief valve piping system. The ANSYS computer program is a large-scale, general purpose computer program for the solution of several classes of engineering analyses. Analysis capabilities include: static and dynamic; elastic and plastic; small and large deflections; linear and non-linear. The matrix displacement method of analyses based upon finite element idealization is used. The library of finite elements includes: elastic pipe, tee, elbow, beam and shell elements; plastic pipe elbow, beam and shell elements; substructure (superelements); spring; mass elements. The loading on the structure may be in the form of forces, displacements, pressures, temperatures or response spectra. ANSYS has been verified and quality assured for Nuclear Safety Related analyses.
For this structural analysis, the straight pipe sections were modeled as elastic pipe elements, the pipe tees were modeled as elastic pipe tee elements, the valves and pipe supports were modeled as explained below in
6 Item 13c, the pipe elbows were modeled as elastic pipe elbow elements except that the three 12 inch diameter elbows in the header at the expected high stress locations were modeled as superelements which were obtained from a detailed elastic shell element model of the 12 inch diameter elbow.
NRC QUESTION 13b 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 ANSYS has the capability to allow the structural analyst to apply the fluid forces directly to the nodal locations of interest (i.e. locations of high stress levels.) As mentioned above in Item 12b and Item 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 l evel s.
As mentioned in Item 12b, REPIPE calculated the wave and blowdown forces for the desired locations and then the force time history was applied to the ANSYS structural model in order to determine the stress levels of the discharge piping system.
NRC QUESTION 13c A description of methods used to model supports, the pressurizer and relief tank connections, and the safety valve bonnet assemblies and PORV actuator.
ODAI RESPONSE Standard structural modeling practices were followed in developing the ANSYS structural model of the discharge piping system. These include the following:
1.
Pipe Supports l
l The mass of the support clamps and the mass of the dynamic portion of the support attached to the pipe were modeled as a lumped mass and placed on the pipe node at or very near to its physical location. The values for the masses were obtained from References 17 and 18. A node at its physical location corresponding to the centerline of the pipe was used to represent the end of the support attached to the pipe. A node at its physical location was used to represent the end of the support not attached to the pipe.
This node was constrained in all degrees of freedom. An ANSYS spring element was used to connect the two nodes of the support. The
\\
l
7 values for the spring constants were obtained from References 19 and 20.
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 from References 17 and 18. The masses and forces were placed on the pipe nodes at or very near to their physical locations.
2.
Pressurizer and Relief Tank Connections 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 freedom.
3.
Valves All of the valves were modeled using three relatively stiff beam elements and a mass element at the valve center of gravity as follows:
one beam 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 beam 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 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 node beam 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 anchor end of the stand at its physical location and was constrained in all degrees of freedom.
NRC QUESTION 13d An identification of the load combinations performed in the analysis together with 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 piping and support adequacy.
e 8
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 Item 12 above for description of this event). Because this type of event is classified as an occasional load, (Level D Service), NB-3656 applies for the seismic Class 1 piping (i.e. piping upstream of the safety / relief valves) and ND-3655 applies for the nonseismic Class D piping (i.e. piping downstream of the safety / relief valves and the header) per Reference 22. That is, the load combination consists of the sum of the sustained loads during normal plant operation and the dynamic load caused by the slug flow. For the nonseismic piping, the peak pressure must be used if the Design Specification states that the peak pressure and slug flow occur simultaneously. 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 be included in the load combination. For this load combination the square root of the sum of the squares (SRSS) method is used.
From NB-3656, the allowable stress for Level D service is 3.0 S, but not greater that 2.0 S,.
While from ND-3655, the allowable stress for Level D service is 3.0 S but not greater than 2.0 S.
h For the materials under consideraiton in this analysis, values of the allowable stress (S ) as a function of temperature are given in Table ?.
NRC QUESTION 13e An evaluation of the results of the structural analysis, including identification of over stressed locations and a description of modifications, if any.
ODAI RESPONSE (Later)
NRC OUESTION 13f A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces.
ODAI RESPONSE iRC Q EST 0 139
9 A copy of the contractors structural analysis report.
ODAI RESPONSE Later.
l l
l I
1 I
1 l
{
10 TABLE 1 Fluid Initial Conditions item (s)
Conditions 1.
Fluid in pressurizer and fluid Saturated steam at the safety upstream of the loop seals valve set point pressure (2499.7 psia) 2.
Fluid in the loop seals See Figure 2 for temperatures 3.
Fluid downstream of the safety Air-water mixture 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 80*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.)
Material Range (*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 Valves 3
40 0.216 SA-312 TP304 110 Downstream of Safety Relief Valves 6
40 0.280 SA-312 TP304 110 Header 12 40 0.406 SA-358 316 80-110 SA-358 316 SA-312 TP304 SA-376 TP316 Temp.
S 1)
S I
A A
A
(*F)
(ksi)
(ksi)
(ksi)
TDT T6 4 T6 T 66 6
{
200 51.6 50.0 51.6 300 46.6 45.0 46.6 400 42.8 41.4 42.8 500 39.8 38.8 39.8 600 37.6 36.4 37.6 650 37.0 35.8 37.0 700 36.2 35.4 36.2 (1) Nonseismic Class D Service, ND-3655 (2) Seismic Class 1 Service, NB-3656
. - _, -...,. _... -. - - ~ - - -., _ - - - _ _.-
12 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 RELAP5/ MODI 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.
Cajigas, J. M., " Verification of the RELAPS-FORCE Hydraulic Force Calculation Code," Gilbert Associates, Inc., May 1984.
7.
Semprucci, L. B. and Holbrook, B.
P., "The Applciation of RELAP4/REPIPE to determine Force Time Histories on Relief Valve Discharge Piping,"
ASME, PVP-33, June 1977.
8.
Strong, B.
R., Jr. and Baschiere, 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.
1
13
- 13. " Criteria and Guidelines for the Design of Safety and Relief Valve Installation in Westinghouse Pressurized Water Reactor Plants,"
Westinghouse Electric Corporation, NES, PWR Systems Division, October 1972.
- 14. Ransom, V. H., et al., 'RELAP5/M001 Code Manual," Volumes 1-2, NUREG/CR-1826, EGG-2070 Draf t, 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:
Hanger No.
Date Hanger No.
Date 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 1RC151-RV1 4-21-77 RCRS-1114 6-02-71 1RC157-RV1 8-25-77 RCRS-1115 11-20-72 l
1RC157-RV2 4-21-77 RCRS-1119 2-16-73 RCH-1005 10-27-72 RCRV-001 l
12-21-72 RCH-1007 1-12-73
- 18. Stone & Webster Bulletin 79-14 Modification Support Drawings:
Hanger No.
Date RCH1006 2-10-81 RCH1010 1-30-81 RCRS1117 1-30-81 RCRS1118 1-30-81 RCRS1120 2-04-81 RCRS1121 1-30-81 RCRS1122 1-30-81 RCRS1123 2-04-81 RCRS1117A 7-22-81 i
l
.9*
14 RCRS1117B 7-22-81 RCRS1118A 7-22-81 RCS1011 RCS1012 RCS1013
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e 15 A
- 19. Books 1 through 6, inclusive, of Stone & Webster, " Zion Station Pipe i
Stress and Support Analysis Report," Number 13430RC - 2, 3, 4, 5, Revision 17, 1983, Commonwealth Edison Job Order 13430.01 for 0, dated January (Pressurizer IRC002 to Pressurizer Relief Tank 1RC003).
- 20. Sargent & Lundy Report No. 037064, Project No. 6320-00.
" Dynamic 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.
. -. -