ML20027E435

Pressurizer Safety & Relief Line Piping & Support Evaluation.

ML20027E435
Person / Time
Site:	Farley
Issue date:	10/31/1982
From:	Cjang K, Siddle D, Laura Smith WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20027E434	List: ML20027E433 ML20027E435
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NUDOCS 8211150268
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WESTINGHOUSE PROPRIETARY CLASS 3 PRESSURIZER SAFETY AND RELIEF LINE PIPING AND SUPPORT EVALUATION ALABAMA POWER COMPANY J. M. FARLEY UNIT 1 AND UNIT 2 l

I L. C. Smith l D. R. Siddle f

October 1982 Approved:

K. C. Chang, Mana Systems Structural %nalysis 1

1 This report is applicable to J. M. Farley Unit 1 and Unit 2 and contains the structural evaluation of ASME III Nuclear Class 1 piping analyzed to requirements of the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components,1971 Edition, including applicable addenda; as well as NNS piping done to requirements of ANSI B31.1 Code, 1967 Edition up to and including 1971 addenda. Results from the Safety and Relief Valve Test program, conducted by the Electric Power Research Institute (EPRI) and concluded on or before July 1,1982, were factored into the analyses presented herein.

8211150268 921104 PDR ADOCK 0S000348 P PDR

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TABLE OF CONTENTS Section Title 1 INTRODUCTION 2 PIPE STRESS CRITERIA 2.1 Pipe Stress Calculation - Class 1 Portion 2.2 Pipe Stress Calculation - Class NNS Portion 2.3 Load Combinations 3 LOADING CONDITIONS ANALYSED 3.1 Loading 3.1.1 Themal Expansion 3.1.2 Pressure 3.1.3 Weight 3.1.4 Seismic 3.1.5 Transients 3.1.6 Safety and Relief Valve Thrust 3.2 Design Conditions 3.2.1 Design Pressure l 3.2.2 Design Temperature 3.3 Plant Operating Conditions 3.3.1 Nomal Conditions 3.3.2 Upset Conditions 3.3.3 Emergency Conditions 3.3.4 Faulted Conditions 0440s:10

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I TABLE OF CONTENTS (Cont)

Section Title 4 ANALYTICAL METHODS AND MODELS 4.1 Introduction 4.2 Static Analysis 4.3 Dynamic Analysis 4.4 Seismic Analysis 4.5 Thermal Transients 4.6 Pressurizer Safety and Relief Line Analysis 4.6.1 Plant Hydraulic Model i- 4.6.2 Comparison to EPRI Test Results 4.6.3 Valve Thrust Analysis l 5 METHOD OF STRESS EVALUATION i

5.1 Introduction 5.2 Primary Stress Evaluation 5.2.1 Design -Conditions 5.2.2 Upset Conditions 5.2.3 Energency Conditions 5.2.4 Faulted Conditions 5.3 Secondary Stress Evaluation 6 RESULTS 6.1 Evaluation Prior to EPRI Test Program 6.2 Evaluation Subsequent to EPRI Test Program 6.2.1 Thermal Hydraulic Results 6.2.2 Structural Results 6.3 Summary of Results and Conclusions l

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, , l SECTION 1 INTRODUCTION The Pressurizer Safety and Relief Yalve (PSARV) discharge piping system for pressurized water reactors, located on the top of the pressurizer, provides overpressure protection for the reactor coolant system. A water seal is maintained upstream of each pressurizer safety and relief valve to prevent a steam interface at the valve seat. This water seal practically elimint.tes the possibility of valve leakage. While this arrangement maximf zes the plant availability, the water slug, driven by high system prersure upon actuation of the valves, generates severe hydraulic shocx loads on the piping and supports.

Under NUREG 0737,Section II.D.1, "Perfomance Testing of BWR and PWR

Relief and Safety Valves", all operating plant licensees and applicants are required to conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents. In addition to the qualification of valves, the functionability and structural integrity of the as-built l discharge piping and supports must also be demonstrated on a plant l specific basis.

In response to these requirements, a program for the performance testing of PWR safety and relief valves was fomulated by EPRI. The primary objective of the Test Program was to provide full scale test data con-fiming the functionability of the reactor coolant system power operated relief valves and safety valves for expected operating and accident l conditions. The second objective of the program was to obtain sufff-cient piping thermal hydraulic load data to pemit confirmation of models which may be utilized for plant unique analysis of safety and relief valve discharge piping systems. Based on the resul'ts of the aforementioned EPRI Safety and Relief Valve Test Program, additional thermal hydraulic analyses are required to adequately define the loads on the piping system due to valve actuation.

This report is the response of the Alabama Power Company to the US NRC plant specific submittal request for piping and support evaluation and is applicable to the J. M. Farley Unit 1 and Unit 2 PSARY piping system.

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1 SECTION 2 PIPE STRESS CRITERIA 2.1 PIPE STRESS CALCULATION - CLASS 1 PORTION In general, the criteria for the structural evaluation of the Class 1 components is based upon two categories of loading. These are self-limiting loads and non-self-limiting loads. A non-self-limiting load produces a primary stress while a self-limiting load produces a secon-dary stress. In order to prevent catastrophic failure of the system, primary stress criteria must be satisfied, which can be accomplished by applying Equation (9) of paragraph NB-3652 of the ASME Boiler and Pressure Vessel Code Section III, up to and including the Summer 1971 Addenda. Fatigue failure may occur if the maximum stress from all loadings is so concentrated at one location that continued cycling of the loads produces a crack, which may then propagate through the wall and result in leakage. For protection against fatigue failure, cyclic stresses from both self-limiting and non-self-limiting loads must be considered. The component will cycle within acceptable limits for each j

specified loading combination if Equation (10), subparagraph NB-3653.1 of the Code is satisfied. This requirement insures that incremental distortion will not occur. The peak stress intensity defined by I

l Equation (11) is then used for calculating the alternating stress intensity, S al t. The value of S alt is then used to calculate the usage factor for the load set under consideration. The cumulative usage factor is then obtained using Miner's rule by considering all other load sets. However, if Equation (10) is not satisfied, which means some plastic defomation occurs with each application of load, the alternate analyt's, " Simplified Elastic-Plastic Discontinuity Analysis", described in subparagraph NB-3653.6 of the Code must be considered. To avoid the posibility of fatigue failure, the cumulative usage factor should not exceed 1.0.

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2.2 PIPE STRESS CALCULATION - CLASS NNS PORTION The piping between the valves and the pressurizer relief tank shall be analyzed to satisfy the requirements of the appropriate equations of the ANSI B31.1 Code. These equations establish limits for stresses from sustained loads and occasional loads (including earthquake), thennal expansion loads, and sustained plus thennal expansion loads, respec-ti vely. The allowable stresses for use with the equations were detennined in accordance with the requirements of the ANSI B31.1 Code.

2.3 LOAD COMBINATIONS In order to evaluate the pressurizer safety and relief valve piping, appropriate load combinations and acceptance criteria were developed.

The load combinations and acceptance criteria are identical to those recommended by the piping subcommittee of the PWR PSARY test program and are outlined in Tables 2-1 and 2-2 with a definition of load abbreviation provided in Table 2-3.

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TABLE 2-1 LOAD COMBINATIONS AM) ACCEPTANCE CRITERIA FOR PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - UPSTREAM OF VALVES Piping Plant / System Allowable Stress Combinatio'n Operating Condition Load Combination Intensity 1 Normal N 1. 5 S, 2 Upset N + OBE + SOT U 1.8 S ,

3 Emergency N + S OT 2.25 S, E

4 Faul ted N + MS/FWPB or DBPB 3.0 S,

+ SSE + SOTp 5 Faul ted N + LOCA + SSE + SOTp 3.0 S, NOTES: ( 1) Plants with an FSAR may use their original design basis in conjunction with the appropriate system operating transient i definitions in Table 2-3; or they may use the proposed criteria contained in Tables 2-1 to 2-3.

( 2) See Table 2-3 for SOT definitions and other load abbreviations.

( 3) The bounding number of valves (and discharge sequence if setpoints are significantly different) for the applicable system operating transient defined in Table 2-3 should be used.

(4) Verification of functional capability is not required, but allowable loads and accelerations for the safety-relief valves must be met.

( 5) Use SRSS for conbining dynamic load responses.

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TABLE 2-2 i

LOAD COMBINATIONS AE ACCEPTANCE CRITERIA I FOR PRESSURIZER SAFETY AND RELIEF VALVE PIPING AND SUPPORTS - SEISMICALLY DESIGNED DOWNSTREAM PORTIGN Piping i

Plant / System Allowable Stress Combination Operating Condition Load Combination Intensity l 1 Normal N 1.0 S h 2 Upset N + S OT U

1. 2 S h 4

I 3 Upset N + OBE + SOT U 1.8 S h 4 Emergency N + S OT E

1.8 S h l

5 Faul ted N + MS/FWPB or DBPB 2. 4 S h

+ SSE + SOTp i

! 6 Faul ted N + LOCA + SSE + SOTp 2. 4 S h j NOTES: (1) Plants with an FSAR may use their original design basis in conjunction with the appropriate system operating transient definitions in Table 2-3; or they may use the proposed

! criteria contained in Tables 2-1 to 2-3.

(2) This table is applicable to the seismically designed portion of downstream non-Category I piping (and supports) necessary

to isolate the Category I portion from the non-seismically l designed piping response, and to assure acceptable valve loading on the discharge nozzle.

l j (3) See Table 2-3 for SOT definitions and other load abbreviations.

i (4) The bounding number of valves (and discharge sequence if setpoints are significantly different) for the applicable system operating transient defined in Table 2-3 should be used.

. (4) Verification of functional capability is not required, but allowable loads and accelerations for the safety-relief valves must be met.

(5) Use SRSS for combining dynamic load responses.

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TABLE 2-3 l DEFINITIONS OF LOAD ABBREVIATIONS N = Sustained loads during nomal plant operation i '9T = System operating transient SOTU = Relief valve discharge transient (1) 2 SOTE = Safety valve discharge transient (1), (2)

SOTF = Max (SOTu; SOT E ); or transition flow OBE = Operating basis earthquake SSE = Safe shutdown earthquake MS/FWPB = Main steam or feedwater pipe break DBPB = Design basis pipe break l

LOCA = Loss-of-coolant accident Sh = Basic material allowable stress at maximum (hot) temperature Sm - Allowable design stress intensity I

(1) May also include transition flow, if detemined that required i

operating procedures cuald lead to this condition.

(2) Although certain nuclear steam supply systems design transients (for example, loss of load) which are classified as upset condi-tions may actuate the safety valves, the extremely low number of actual safety valve actuations in operating pressurizer water reactors justifies the emergency condition from the ASME design .

philosophy and a stress analysis viewpoint. However, if actuation of safety valves would occur, a limitation must be placed to shut down the plant for examination of system integrity after an appro-priate number of actuations. This number can be detemined on a plant specific basis.

NOTE: Plants with an FSAR may use their original design basis in conjunction with the appropriate system operating transient definitions in Table 2-3; or they may use the proposed criteria contained in Tables 2-1 to 2-3.

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SECTION 3 LOADING CONDITIONS !.NALYZED 3.1' LOADING The piping stress analyses described in this section consider the load-

ings specified in the design specification. These loadings result from

! themal expansion, pressure, weight, earthquake, design basis accident (DBA), plant operational themal and pressure transients, and safety valve and relief valve operation.

3.1.1 THERMAL EXPANSION l

The thermal growth of the reactor coolant loop equipment and all connected piping is considered in the themal analysis of this system.

l The modulus of elasticity, (E), the coefficient of thermal expansion at l the metal temperature, (a), the external movements transmitted to the I

piping as described above, and the temperature rise above the ambient l temperature, (AT), define the required input data to perfom the flexi-bility analysis for thermal expansion.

Due to different operating modes, the system may experience multiple themal loadings. The temperatures used in the expansion analysis of the piping are based upon the infomation presented in the design documents.

3.1.2 PRESSURE Pressure loading in this report is either design pressure or operating pressure. The design pressure is used in the calculation of longitu-dinal pressure stress in accordance with the Code. The range of oper- ,

ating pressure is used in calculating various stress intensities, as applicable.

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3.1.3 WEIGHT To meet the requirements of the Code, a weight analysis is perfomed by applying a 1.0 g unifomly distributed load downward on the complete piping system. The distributed weight characteristics of the piping system are specified as a function of its properties. This method

provides a distributed loading to the piping system as a function of the weight of the pipe, insulation, and contained fluid during nomal oper-ating conditions.

3.1.4 SEISMIC l Seismic motion of the earth is treated as a random process. Certain assumptions reflecting the characteristics of t;ypical earthquakes are j made so these characteristics can be readily employed in a dynamic response spectrum analysis.

Piping rarely experiences the actual seismic motiva at ground elevation, since it is supported by components attached to the containment build-ing. Although a band of frequencies is associated with the ground earthquake motion, the building itself acts as a filter to this environ-ment and will effectively transmit those frequencies corresponding to its own natural modes of vibration.

The forcing functions for the piping seismic analyses are derived from dynamic response analyses of the containment building when subjected to seismic ground motion. These forcing functions are in the form of floor response spectra. Response spectra are obtained by detemining the maximum response of a .iingle mas... spring-damper oscillator to a base motion time history. This single mass-spring-damper oscillator system represents a single natural vibration mode of the piping system. A plot of the maximum responses versus the natural frequencies of the oscil-lator foms the response spectrum for that particular base motion.

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The intensity and character of the earthquake motion producing forced vibration of the equipment mounted within the containment building are specified in terms of the floor response spectrum curves at various elevations within the containment building.

The seismic floor response spectrum curves corresponding to the highest elevation at which the component or piping is attached to the contain-ment building are used in the piping analysis.

Seismic loads must be known to calculate the resultant moment (Mi3 )

used in the design equations The plant operating condition (full load) is the condition under which the specified earthquake is assumed to occur.

3.1.5 TRANSIENTS To provide the necessary high degree of integrity for the NSSS, the transient conditions selected for secondary stress evaluation are based on conservative estimates of the magnitude and anticipated frequency of occurrence of the temperature and pressure transients resulting from the I possible operating conditions.

I The transients selected are conservative representations of transients '

for design purposes, and are used as a basis for piping secondary stress evaluation to provide assurance that the piping is acceptable for its application over the design life of the plant.

For purposes of piping evaluation, the number of transient occurrences are based on a plant design life of 40 years.

3.1.6 SAFETY AND RELIEF VALVE THRUST The pressurizer safety and relief valve discharge piping system provide overpressure protection for the RCS. The three spring-loaded safety 0440s:10

r valves and two power-operated relief valves, located or, top of the pressurizer, are designed to prevent system pressure from exceeding design pressure hy more than 10 percent and 100 psi, respectively. A water seal is maintained upstream of each valve to minimize leakage.

Condensate accumulation on the inlet side of each valve prevents any leakage of hydrogen gas or steam through the valves. The valve outlet side is sloped to prevent the formation of additional water pockets.

If the pressure exceeds the set point and the valves open, the water slug from the loop seal discharges. The water slug, driven by high system pressure, generates transient thrust forces at each location where a change in flow direction occurs.

The safety and relief lines are analyzed for various cases of thrust loadings to ensure the primary and secondary stress limits are not exceeded.

1 3.2 DESIGN CONDITIONS l

The design conditions are the pressures, temperatures, and various mechanical loads applicable to the design of nuclear power plant piping.

3.2.1 DESIGN PRESSURE The specified internal and external design pressures are not less than the maximum difference in pressure between the inside and outside of the component, which exists under the specified normal operating condi-tions. The design pressures are used in the computations made to show l compliance with the Code (subparagraph 101.20 of the Code).

3.2.2 DESIGN TEMPERATURE l The specified design temperature is not less than the actual maximum l

l metal temperature existing under the specified normal operating condi-l 0440s:10

tions for each area of the component considered. It is used in computa-tions involving the design pressure and coincidental design mechanical loads (subparagraph 101.3 of the Code).

3.3 PLANT OPERATING CONDITIONS 3.3.1 NORMAL CONDITIONS A normal condition is any condition in the course of system startup, design power range operation, hot standby, and system shutdown, other than upset, faulted, emergency, or testing conditions.

i 3.3.2 UPSET CONDITIONS 4

An upset condition is any deviation from normal conditions anticipated to occu often enough that design should include a capability to with-stand the condition without operational impairment. Upset conditions include those transients resulting from any single operator error or

control malfunction, transients caused by a fault in a system component

( requiring its isolation from the system, and transients due to loss of loao or power. U, set conditions include any abnomal incidents not resulting in a forced outage and also forced outages for which the corrective action does not include any repair of mechanical damage.

3.3.3 EMERGENCY CONDITIONS Emergency conditions are defined as those deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of any damage developed in the system. The total number of postulated occur-rences for such events shall not cause more than 25 stress cycles (subparagraph NB-3113.3 of the code).

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3.3.4 FAULTED CONDITIONS Faulted conditions are those combinations of conditions associated with extremely low probability - postulated events whose consequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent that considerations of public health and safety are involved.

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SECTION 4 l ANALYTICAL METHODS AND MODELS

4.1 INTRODUCTION

I The analytical methods used to obtain a piping deflection solution consist of the transfer matrix method and stiffness matrix fonnulation for the static structural analysis. The response spectrum nethod is used for the seismic dynamic analysis.

l The complexity of the piping system requires the use of a computer to obtain the displacements, forces, and stresses in the piping and support i

members. To obtain these results, accurate and adequate mathematical

! representations (analytical models) of the systems are required. The l

modeling considerations depend upon the degree of accuracy desired and the manner in which the results will subsequently be interpreted and evaluated. All static and dynamic analyses are performed using the WESTDYN computer program. This program, described in WCAP-8252, was reviewed and approved by the U.S. NRC (NRC letter, Aprif 7,1981 from i R. L. Tedesco to T. M. Anderson).

The integrated piping / supports system model is the basic system model used to compute loadings on components, component and piping supports, and piping. The system model includes the stiffness and mass charac-teristics of the piping, attached equipment, and the stiffness of l

supports, which affects the system response. The deflection solution of l the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.

4.2 STATIC ANALYSIS The piping system models, constructed for the WESTDYN computer program, are represented by an ordered set of data, which numerically describes the physical system.

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The spatial geometric description of the piping model is based upon the isometric piping drawings referenced in this report and equipment draw-ings referenced in the design specification. Node point coordinates and incremental lengths of the members are detemined from these drawings.

Node point coordinates are put on network cards. Incremental member lengths are put on element cards. The geometrical properties along with the modulus of elasticity, E, the coefficient of themal, expansion, a, the average temperature change from the ambient temperature, AT, and the weight per unit length, w, are specified for each element. The supports are represented by stiffness matrices which define restraint character-istics of the supports. Plotted models for various parts of the safety and relief valve discharge piping are showr, in figures in Section 6.

The static solutions for deadweight and thermal loading conditions are obtained by using the WESTDYN computer program. The WESTDYN computer program is based on the use of transfer matrices which relate a twelve-element vector [B] consisting of deflections (three displacements and three rotations) and loads (three forces and three moments) at one loca-tion to a similar vector at another location. The fundamental transfer matrix for an element is determined from its geometric and elastic prop-erties. If themal effects and boundary forces am included, a modified transfer relationship is defined as follows: 1 T T A a A il 12 o t i

+ =

T T F, f F

-21 22- _ _ _.

t 4

l or l

TBgo+Ry=By where the T matrix is the fundamental transfer matrix as described above, and the R vector includes themal effects and body forces. This B vector for the element is a function of geometry, temperature, coeffi-l l

cient of thermal expansion, weight per unit length, lumped masses, and externally applied loads.

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The overall transfer relationship for a series of elements (a section) can be written as follows:

By TB1o+R1 B2=T021+R2-TTB2yo+TR2g+R2 B3 TB32+R 3=TTTB321o+TTR3 2 y +.T R32+R3 B

n" l fn \'

J o

+

n E '

h w T R r-1 +R n

"-2

( r) (" /r l A network model is made up of a number of sections, each having an over-all transfer relationship formed from its group of elements. The linear elastic properties of a section are used to define the characteristic i stiffness matrix for the section. Using the transfer relationship for a section, the loads required to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the section are obtained. These loads are incorporated in the overall load vector.

After all the sections have been defined in this manner, the overall stiffness matrix, X, and associated load vector needed to suppress the deflection of all the network points is determined. By inverting the l stiffness matrix, the flexibility matrix is detemined. The flexibility matrix is multiplied by the negative of the load vector to determine the network point deflections due to the themal and boundary force effects. Using the general transfer relationship, the deflections and internal forces are then detemined at all node points in the system.

The support loads, F, are also computed by multiplying the stiffness matrix, K, by the displacement vector, 6, at the support point.

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4.3 DYNAMIC ANALYSIS The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment.

4.4 SEISMIC ANALYSIS The lumping of the distributed mass of the piping systems is accom- ,

plished by locating the total mass at points in the system which will appropriately represent the response of the distributed system. Effects of the equipment motion, that is, the pressurizer, on the piping system are obtained by modeling the mass and the stiffness characteristics of the equipment in the overall system model.

I The supports are anain represented by stiffness matrices in the system model for the dynamic analysis. Mechanical shock suppressors which resist rapid motions are now considered in the analysis. The solution for the seismic disturbance employs the response spectra method. This method employs the lumped mass technique, linear elastic properties, and the principle of modal superposition.

l l From the mathematical description of the system, an overall stiffness matrix [K] is developed from the individual element stiffness matrices using the transfer matrix [K ] Rassociated with mass degrees-of-freedom only. From the mass matrix and the reduced stiffness matrix, the natural frequencies and the normal modes are detennined. The modal participation factor matrix is computed and combined with the appro-priate response spectra value to give the modal amplitude for each mode. Since the modal amplitude is shock direction dependent, the total i modal amplitude is obtained conservatively by the absolute sum of the contributions for each direction of shock. The modal amplitudes are then converted to displacements in the global coordinate system and applied to the corresponding mass point. From these data the forces, I

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r-moments, deflections, rotation, support reactions, and piping stresses are calculated for all significant modes.

The seismic response from each earthquake component is computed by combining the contributions of the significant modes.

4.5 THERMAL TRANSIENTS Operation of a nuclear power plant causes temperature and/or pressure fluctuations in the fluid of the piping system. The transients for this system are defined in " Westinghouse Systems Standard Design Criteria 1.3" and referenced in the Design Specification and were used to define the various operating modes used in the thermal expansion analyses.

, 4.6 PRESSURIZER SAFETY AND RELIEF LINE ANALYSIS 4.6.1 PLANT HYDRAULIC MODEL When the pressurizer pressure reaches"the set pressure (2,500 psia for a safety valve and 2,350 psia for a relief valve) and the valve opens, the high pressure steam in the pressurizer forces the water in the water

! seal loop through the valve and down the piping system to the pressurizer relief tank. For the pressurizer safety and relief piping system, analytical hydraulic models, as shown in Figures 4-1 and 4-2, were developed to represent the conditions described above.

The computer code ITCHVALVE was used to perfom the transient hydraulic analysis for the system. This program uses the Method of Characteris-tics approach to generate fluid parameters as a function of time. One-dimensional fluid flow calculations applying both the implicit and explicit characteristic methods are perfomed. Using this approach the l piping network is input as a series of single pipes. The network is generally joined together at one or more pl' aces by two or three-way junctions. Each of the single pipes has associated with it friction factors, angles of elevation and flow areas.

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l Conservation equations can be converted to the following characterisitic equations:

h=V+c q' c h + pc h = c(F + pgcose) h

• W h=V-c h - oc h = -c(F + pgcose) 4

'W c2 , - ah/ap ah 1 5 ~ pI z = variable of length measurement t = time Y = fluid velocity c = sonic velocity p = pressure p = fluid density l

F = flow resistance g = gravity

! e = angle off vertical 1

i J = conversion factor for converting pressure units to j equivalent heat units h = enthalpy

! q = rate of heat generation per unit pipe length The computer program possesses special provisions to allow analysis of valve opening and closing situations.

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m t

Fluid acceleration inside the pipe generates reaction forces on all segments of the line that are bounded at either end by an elbow or bend. Reaction forces resulting from fluid pressure and momentum variations are calculated. These forces can be expressed in tems of the fluid properties available from the transient hydraulic analysis, l perfomed using program ITCHVALVE. The momentum equation can be expressed in vector form as:

F =

pVdv + h pV(V

• ndA) ev c t jv J From this equation, the total force on the pipe can be derived:

r y (1 - cos alI aW r 2 (1 - cos a2} aW pipe *{ sin at af Bend 1 9c sin a2 at Bend 2 straight hdi

+hcjpipe A = piping flow area y = volume F = force r = radius of curvature of appropriate elbow a = angle of appropriate elbow W = mass acceleration All other tems are previously defined.

Unbalanced forces are calculated for each straight segment of pipe from l the pressurizer to the relief tank using program FORFUN. The time-histories of these forces are stored on tape to be used for the subse-quent structural analysis of the pressurizer safety and relief lines.

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4.6.2 COMPARISON TO EPRI TEST RESULTS Piping load data has been generated from the tests conducted by EPRI at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative commercial upstream environments were carried out. The resulting downstream piping loadings and responses were measured. Upstream environments for

particular valve opening cases of importance, which envelope the commercial scenarios, are

A. Cold water discharge followed by steam - steam between the pressure source and the loop seal - cold loop seal between the steam and the

! val ve, B. Hot water discharge followed by steam - steam between the pressure l source and the loop seal - hot loop seal between the steam and the

" val ve.

1

! I C. Steam discharge - steam between the pressure source and the valve,  !

l Specific thermal hydraulic and structural analyses have been completed for the Combustion Engineering Test Configuration. Figure 4-3 illus-trates the placement of force measurementJensors at the test site.

Figures 4-4, 4-5 and 4-6 illustrate a comparison of the thermal hydrau-lically calculated results using the ITCHVALVE and FORFUN computer programs versus experimental results for Test 908, the cold water discharge followed by steam case. Figure 4-4 shows the pressure time histories for PT9, which is located just downstream of the valve.

Figures 4-5 and 4-6 illustrate, respectively, the force time histories of the horizontal run (WE28/WE29) and the long vertical run (WE32/WE33) imediately downstream of the safety valve. Significant structural damping in the third segment after the valve was noticed at the test and was verified by structural analyses. Consequently, a comparison of force WE30/WE31 was not presented here. No useable test data for sensor WE34/WE35 was available for Test 908.

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Figures 4-7 through 4-11 illustrate a comparison of calculated versus experimental results for Test 917, the hot water discharge followed by steam case. Figure 4-7 shows the pressure time histories for PT9.

Figures 4-8, 4-9, 4-10 and 4-11 illustrate, respectively, the themal hydraulically calculated and the experimentally detemined force time histories for (WE28/WE29), (WE32/WE33), (WE30/WE31) and (WE34/WE35).

Blowdown forces were included in the total analytically calculated force for WE34/WE35 as this section of piping vents to the atmosphere.

Although not presented here, comparisons were also made to the test data available for safety valve discharge without a loop seal (steam discharge).

The application of the ITCHVALVE and FORFUN computer programs for cal-l culating the fluid-induced loads on the piping downstream of the safety

^

and relief valves has been demonstrated. Although not presented here, i the capability has also been shown by direct ccmparison to the solutions of classical problems.

The application of the structural computer programs (discussed in Section 4.6.3) for calculating the system response has also been demonstrated. Structural models representative of the Combustion Engineering Test Configuration were developed. Figures 4-12, 4-13 and 4-14 illustrate, respectively, a comparison of the structural analysis l

results and the experimental results for locations (WE28/WE29),

(WE32/WE33) and (WE30/WE31) for test 908. No useable test data for sensor (WE34/WE35) was available. Figures 4-15, 4-16, 4-17 and 4-18 l

l show for test 917, respectively, the structural analysis results versus l the test results for locations (WE28/WE29), (WE32/WE33), (WE30/WE31) and

! (WE34/WE35).

4.6.3 YALVE THRUST ANALYSIS The safety and relief lines were modeled statically and dynamically.

(seismically) as described in Sections 4.1 through 4.4. The mathe-matical model used in the seismic analysis was modified for the valve thrust analysis to represent the safety and relief valve discharge. The 1

0440s:10

time-history hydraulic forces determined by FORFUN were applied to the piping system lump mass points. The dynamic solution for the valye thrust was obtained by using a modified-predictor-corrector-integration technique and normal mode theory.

The time-history solution was found using program FIXFM3. The input to this program consists of natural frequencies, nomal modes, and applied forces. The natural frequencies and normal modes for the modified pres-surizer safety and relief line dynamic model were detemined with the WESTDYN program. The time-history displacement response was stored on magnetic tape for later use in computing the total system response due to the valve thrust conditions. The time-history displacements of the FIXFM3 program were used as input to the WESTDYN2 program to detemine the time-history internal forces and deflections at each end of the piping elements. For this calculation, the displacements were treated as imposed deflections on the pressurizer safety and relief line masses. The solution was stored on tape for later use in the piping stress evaluation and piping support load evaluation.

The time-history internal forces and displacements of the WESTDYN2 program were used as input to the POSDYN2 program to detemine the maximum forces, moments, and displacements that exist at each end of the piping elements and the maximum loads for piping supports. The results from program POSDYN2 are saved on TAPE 14 for future use in piping stress analysis and support load evaluation.

0440s:10

- _ _ _ _ _ _ _ - _ . _ 1

~

Safety Line B QV031B Safety Line C b

QV031A Pressurizer g

QV031C j Safety 6 Line A 9 2 1 5 3 /

3 7

14

\ l l

11 l 17 Common Header l

4 From ..

Relief ina Pressurizer Relief Tank i

FIGURE 4-1 SAFETY LINES HYDRAULIC MODEL NOTE: The numbers correspond to the force location in Table 6-1.

l

F V027A V053 Valve Set "A" 3

Valve Set QV027B 2 h .

061 13 6

1 pe...ue;sse Q

i From Safety Line C From SafetyLine B

From Safety Line A

8 7 11

=F Pressurisee hti.f T..w FIGURE 4-2 RELIEF LINF HYDRAULIC MODEL NOTE: The numbers correspond to the force locations t Table 6-2.

EY y '

. 71 Se t 'sX / I L WE28 p 2

prog L+

/ + x' WE29 p

M / . Z v -

DISPLACEMENT

} '

F. FORCE /

1 II I

X Segment [

' rw i

' -*- /

1

/

/

/

i WE30 WE31 g

! I i

! PT10 r

/

Segment X

.Se nt

/ 4 A

1. 7x Sy 7-l

--* 3X Fy F '

Y ' PE34

'$ Y O r r r r r r r r r'< <

--Ni t t i st s'rr r rrrr r r rrrrrrr r r rrrrrrr r r r r rr rri s t rrr i FIGURE 4-3:

STRIICTIIRAL RESPONSE - FORCE WASIIDEMENT LnCATIONS - EPRI TESTS

! ~,<

500."

< A

\

l \

400. ,

i i tests i 1 1 1 ---

ITCHVALVE

. , I i i t I

\

l \

O 300,- -

t t C

~ g \

r 1 -

E g i I  !

E

\

' l

\ /

200.- -

,' \ f

\

o s \.//

l 1

I i

100,- - l l

e I

_ _ _ _ l' .

D. Q.1 0.2 0.3 b time (seconds)

FIGURE 4-4 : Comparison of the EPRI Presst:re Time-History for PT09 from Test 908 with the ITCHVALVE Pre-i I

dicted Pressure Time-History

~

1.0E4 A

N\

l I ,

I l <

/'s f/ '

i

' I 0.0 , --

' -~'

j

, I -

l I I I I

I i

- I I I l

$ -1.0E4 '

,l-

_ i w I I I

M i l

\ e \ u

\

\

\

\ \

-2.0E4 h ,

tests ITCHVALVE l

1

-3.0E4 0 05 0.15 0.25 Time (seconds)

FIGURE 4-5: COMPARISON OF THE EPRI FORCE TIME-HISTORY FOR WE28 and WE29 FROM TEST 908 WITH THE ITCHVALVE PREDICTED FORCE' TIME-HISTORY

i 1.0E5 1

g

=

0

' )I $Vf f e

E

}

-1.0ES. .

-2.0E5 l 1

l tests

- ITCEVALVE

-3.0E5 0.1 0.2 0.3 0.4 0.5 time (secondsl FIGURE 4-6: COMPARISON OF THE EPRI FORCE TIME-HISTORY FOR WE12 AND WE33 FROM TEST 908 WITH THE ITCHVALVE PREDICTED FORCE TIME-HISTORY

9 500.

N

\

\

\

400s . I \

g

\

\

\

\

\

\

\

_ 300 -

\

g \

a f \

l \

e \

g I .s .

, g _

I s\ %___

I l ~~~

[ 200. . I I

I I

I Test I

i '---

100 . p ITCHVALVE I

l l

1

)

. _, / % '

!  !  ! l

0. 0.1 0.2 0.3 0.4 0.5 time (seconds)

FIGURE 4-7 : Comparison of the EPRI Pressure Time-History from PT09 from Test 917 with the ITCHVALVE Predicted ,

Pressure Time-History

t 4000 l:

t 2000 0 1 i ., \  : 01

- \\ \

y Al oo a *l \ l,'

1 \ /

l .8 \ ,  ;  ?

\\

t i t

' -2000 g

i i

1

-4000  ;

ll v

-6000 tests

---

ITCHVALVE

-8000 --

0.0 0.1 0.2 0.3 0.4 0.5 0. 6 time (seconds)

FIGURE 48: Comparison of the EPRI Force Time-History for WE28 and WE29 from Test 917 with the ITCHVALVE Predicted Force Time-History l

I 4

j l

l 2.0E4 A

!\

l.0E4 ,I i

' i  !

I

/{i s

\

\, ol \

- , s

$ 0. ,v

!! I r$ A M nvva i

~ v  ; ,y v v 8

k ')\ l i

i i

i i

-1.0E4 ,

I

'sd tests ITCHVALVE

^^

-2.0E4 0.0 0.1 0.2 0.3 0.4 0.5 time (seconds)

FIGURE 4-9 : Comparison of the EPRI Force Time-History for WE32 and WE33 from Test 917 with the ITCHVALVE Predicted Force Time-History

? :E4 -

tests ITCHVALVE.

2.0E4 ,'s

\

\

al \

l \

l t I t I \

q 1.0E4 ,

t a  ! hi's

=

8 l \

i R

s-

)

if 0.

-w ---

> (t'; O s /

\ /

l

( s i -1.0E4 ,/

d

-2.0E4 0.0 0.1 0.2 0.3 0.4 0.5 l time (seconds)

FIGURE 4-10: Comparison of the EPRI Force Time-History For WE30 and WE31 From Test 917 with the ITCHVALVE Predicted Force Time-History l -

l 1

l l

l 2.0E4-tests

- ---- ITCHVALVE I s C I \

y l \

=

! \

e / \

if 1.0E4 l s

\

i t

) \

! \

l I

l

'v ~

l I I

I 0.0 .

_/

0.0 0.1 0.2 0.3 0.4 0.5 time (seconds)

FIGURE 4-11: Comparison of the EPRI Force Time-History For WE34 and WE35 from Test 917 with the ITCHVALVE Predicted Force Time-History 1

. ~ . .

20,0 i

10.0 f

O

'\ \

\ W s) g E 0,0 - / !/j I['4'yI A AO

'v ~<'v-v--

i l i;;

l 8 l I

8. VI

u. I

-10,0 i \, l (l j y Tests

---

FIXFM3

-20.0 (Structural Analysis)

I l II d . -

-26. l e_

0.05 0.15 0.25 0.35 0.45 l

Time (Sec.)

FIGURE 4-12: Comparison of the EPRI Force Time-History for WE28 and WE29 from Test 908 with the FIXFM3 Predicted Force Time-History l

l

f.

• 6 111.01 100.0 '

si 50.0

Ill A l kV, 0.0 ,

~ ,

N ,'

kw l\ r A.,. .

., i l _

50.0 i m I k I4 5 i i

y -

100.0 -

l 2

! -150.0 i

Tests

-200.0 -

'fl

---

FIXFM3 i i

(Structural Ana. lysis) '

246.92 '

0.0 O.1 0.2 0.3 0.4 Time (SEC)

Figure 4-13: Comparison of the EPRI Force Time-History For WE32 and WE33 From Test 908 With the FIXFM3 Predicted Force Time-History

,=w.-w., 4,, , n- --,--.-----,--n. - - , , , .n.,-,, n-- - , , - , - , + - - - - - - , s-- - - - - -- - - - - - - - _ - - - - - - - - - - - - - - - -

90.896 -

0 I I 75.0 '

I l I ,l I I 50.0 l

i s i 3 l I l i

'ql i i e 25.0 l b s n i

!M/I lj I 8

i  :

I i fI a l l l l 'l' I I 0.0 . s  !  :

S i E vv ' i' i pI g

g i,I i

, i \ I I E

I M

l '

if -25.0 ,'

8 i .'! .

n i

l I i i

i f. I i*

g l l I

'I I 8  !

-50.0 ~

Y I l

Tests l!

l1 8

l -75.0 -

3

---

FIXFM3 l#

(Structural Analysis) I#

-98.324 - - 'd 0.0 0.1 0.2 0.3 0.4 Time (SEC)

Figure 4-14: Comparison of the EPRI Force Time-History For WE30 l

and WE31 From Test 908 With the FIXFM3 Predicted Force Time-History 6

1 I

- - - - - - ,---,,--..,--..n - , - - - - - _ w- -~-~,n- w- .,------,._,mw,--,a --. . , . ,,w - - - , - - - - , ,----,-,--,,.,.,em, - - , - . . - - , - - - - - - - - -

5.0 4.0 F

il 3.0 .'

i ,

I l1. '

t 2.0

,f ,

l 'A, t i 1.0  ;

/ h '

i j

! !a W N

0.0 t'

lr n y

^ ~'(--v  ;

' ^ ~ '

'8 I i 1.0

!i !h ! l I i I l h

. I r E - I l j l i

S -2.0 -

ti i i 8 8 5  ! I g -3.0  ;

' )

I i

-4.0 l l

1 l

-5.0 0

-6.0 Tests -7.0

---

FIXFM3

-8.0 -

0.0 0.1 0.2 0.3 0.4 0.495 Time (SEC)

Figure 4-15fomparison of the EPRI Force Time-History For WE28 and WE29 From Test 917 With the FIXFM3 Predicted Force Time-History l

l l

i 1- --. _ -- - . . - . . - _ _ - . - - - - - . - - - - - . - . - - - - - - - - - - - - - - - - - - - - - - - - - . - - - - - - - - - - - - - -

7-I

• l l

12.956 Ii 10.0 fI d g}  ;

!' i i

,! \

, i 5.0 I

yy h! \

j l

! j l I f

,I e I I{ i i I 2 ,

I I I

' b 0'0

- -' ' " ^ '

i II V l,, 'y " y I e b gj lI

(

-5.0 ) /

t 7

/

l / Tests 1

I1 I

---

FIXFM3 f (Structural Analysis)

-10.0 ,

-13.266 0.0 0.1 0.2 0.3 0.4 0.495 Time (SEC)

Figure 4-16: Comparison of the EPRI Force Time-History For WE32 and WE33 From Test 917 With the FIXFM3 Predicted Force Time-History

q- . . . . .

c- .

. 6 25.863 25.0  ?

I ,s II

20.0  :

I h I I 15.0' In \

i.

i

\

10.0 i Ys t ft L1 )

5.0 i  ! l\  ;

, g i t a i s b ..) t h a. i fj -,

g 0.0 y y yyy - -

r y rv\ 1 7

o '

1 y \

/

o  ! /

' 5. 0 l

f

' /

v\

l

-10.0 g

\s' Tests

-15.0 -

---

FIXFM3 Y (Structural Anal-

^ysis) ^

-20.0 0.0 0.1 0.2 0.3 0.4 0.495 Time (SEC)

Figure 4-17: Comparison of. the EPRI Force Time-History For WE30 and WE31 From Test 917 With the FIXFM3 Predicted Force Time-History

l- . ,

• 14.58 g

[ft\

t \

12.5 l t i ~'

I \\

l

,! \.

10.0 i

i I

7.5 E g \

& I

! ~

u I i '

f

$ 5.0 a h '

o f /

Tests i

---

FIXFM3 (Structural Analys ts).

2.5 '

I I

0.0 . .. l 0.0 0.1 0.2 0.3 0.4 0.495 Time (SEC) i Figure 4-18: comparison of the EPRI Force Time-History For i

WE34 and WE35 From Test 917 With the FIXFM3

~

Predicted Force Time-History l

l l

I l

SECTION 5 METHOD OF STRESS EVALUATION

5.1 INTRODUCTION

The method used to combine the primary loads to evaluate the adequacy of the piping system is described in this section.

5.2 PRIMARY STRESS EVALUATION In order to perform a primary stress evaluation in accordance with the rules of the Code, definitions of stress combinations are required for the noma 1, upset, emergency and faulted plant conditions as defined in Section 3. Tables 2-1 and 2-2 illustrate the allowable stress inten-sities for the appropriate combination. Table 2-3 defines all pertinent tems.

5.2.1 DESIGN CONDITIONS The piping minimum wall thickness, mt , is calculated in accordance with the Code. The actual pipe minimum wall thickness meets the Code requirement.

The combined stresses due to primary loadings of pressure, weight, and design mechanical loads calculated using applicable stress intensity factors must not exceed the allowable limit. The resultant moment, Mj , due to loads caused by weight and design mechanical loads is calculated using the following equation:

2+ 2 M + + M

[M*wt [(M I= M*DMV #wt JDM

\

I 2 1/2

+ j M + M z

(zwt DM 0440s:10

f-where M ,M ,M z = deadweight moment components x

wt #wt wt

,M = design mechanical load moment components M*DML,M#DML z DML The maximum stresses due to pressure, weight, and DML in the piping system are reported on tables in Section 6.

5.2.2 UPSET CONDITIONS The combined stresses due to the primary loadings of pressure, weight, OBE seismic, and relief valve thrust loadings calculated using the applicable stress intensity factors must not exceed the allowables. The resultant moments, Mt . due to loads caused by these loadings are calculated as shown below.

For seismic and relief valve thrust loading:

[ 1/2h 2 [ /2h M M + 2 + . g , (g 2 +g 2 l2 I= i (M*0BE M*2 1

_ (*wt OT /

U \ 'vt YOBE Y OTU)

[ 2 2 /2 h 2 1/2

+i M + M I

+(M 0BE 2 z

( z'vt OT U) _

where M ,M ,M = deadweight moment components x z wt #wt wt

,M 2 = inertial OBE moment components M*0BE,M#0BE 0BE M ,M ,M 2 = relief line operation moment components x

SOT # SOT SOT U U U 0440s:10

5.2.3 EMERGENCY CONDITIONS The combined stresses due to primary loadings of pressure, weight and safety valve thrust, using applicable stress intensification factors, must not exceed the allowable limits. The magnitude of the resultant moment, M, is calculated from the moment components as shown below:

f \2 f +

\2 I +

\2 1/2 M M + M M M t M I 4= l j +1 1 z

(

• SOT E wt ) (# SOT g

. #wt) (M* SOT -

wt/

where M ,M ,M = deadweight moment components x

wt #wt wt

,M ,M = safety line operation moment components M* SOT E SOT E

SOT E

5.2.4 FAULTED CONDITIONS l The combined stresses due to primary loadings of pressure, weight, SSE and S0T p , using applicable stress intensification factors must not exceed the allowable limits. For the resultant moment loading, Mj ,

the SSE and SOTp moments are combined using the square-root-of-the-sum-of-the-squares (SRSS) addition and added absolutely with deadweight for each moment component (M x

,M y , Mz ). The magnitude of the resultant moment, Mg , is calculated from the three moment components, l as shown below:

l M

2 . 2$ 1/2 ', 2 i= <

[M*

\ pSOT g*SSE

/ g*wt

~

2 2 1/2 2

+

[M#

\ SOT p

+ M

1. SSE

\

/

+ M

1. wt .

0440s:10

2 2 1/2 2 1/2

+ + I + M )

i. (M* pSOTM*SSEj z wt ,

a where M *N * = deadweight moment components x z wt #wt wt

,M = inertial SSE moment components M*SSE,M#SSE z SSE

= maximum of SOTU or SOTE moment components M* SOT ,M p # SOT , M*S0T p p For the safety and relief piping, the faulted condition load combination of pressure, weight, and valve thrust is considered as given in Tables 2-1 and 2-2 and defined in Table 2-3. The pipe break loads (MS/FWPB or LOCA) can be ignored for the PSARV system. These loads have very little impact on the pressurizer safety and relief system when compared to the loading conditions discussed in this report.

l 5.3 SECONDARY STRESS EVALUATION j The combined stresses due to the secondary loadings of thermal, pres-sure, and deadweight using applicable stress intensification factors must not exceed the allowable limit. For the resultant moment loading, M9 , thermal moments are combined as shown below:

,g 2 ,g 1/2 l M I= }2 .

[# . [(gMAX z z }2

[(M* MAX _M* \ MAXMIN YMI/ MIN /

M ,M ,M = maximum themal moment considering all themal cases MAX YMX zMX including nomal operation l

0440s:10

M ,M ,M = minimum themal moment considering all themal cases MIN JMIN zggy including normal operation This, Mj , is then substituted into the appropriate equations of the applicable code.

l l

I -

l l

l l

l 0440s:10

SECTION 6 RESULTS 6.1 EVALUATION PRIOR TO EPRI TEST PROGRAM The J. M. Farley Unit 1 and Unit 2 safety and relief viuve discharge piping system has received a very detailed themal hydraulic and structural dynamic evaluation to insure the operability and structural integrity of the as-built system (WCAP-9718). This structural eval-uation, including the themal hydraulic analysis, was based on the criteria and methods that were current prior to the availability of the  ;

data from the EPRI Test Program. The thermal hydraulic forcing func-tions were generated assuming simultaneous opening of either the safety valves or the relief valves, since they represent the worst applicable l

loading conditions for the piping and supports for this specific lay-out. These forcing functions were then used as input to the structural I evaluation in which the primary and secondary stresses were detemined.

The methods used and the loadings considered are consistent with Section 2.0 and Section 3.0 of this report, respectively. Results of this extensive analysis and evaluation have demonstrated that the PSARV piping meets all the applicable design limits for the various loading cases. In addition, the acceptability of the valve nozzles, equipment nozzles, and pressurizer shell was assured for the applied loads.

6.2 EVALUATION SUBSEQUENT TO EPRI TESTING The evaluation subsequent to the EPRI Testing Program uses the same procedure as the prior as-built analysis. The only difference occurs in the themal hydraulic evaluation and analysis which uses computer programs which have been shown to match the test results of the EPRI Program. The thermal hydraulic forcing furictions were generated using the same criteria as before, that is, the simultaneous opening of either the safety valves or the relief valves. These forcing functions were input into the structural analysis using the same mathematical model as used in the as-built analysis. The methods used and the loadings considered are consistent with Sections 2.0, 3.0, 4.0 and 5.0 of 0440s:10

- - _ _ - - _ _ . _ . - . _ . _ - - - - . . . - . - - - . - - . . .._ _ _= ._

this report. For the loads other than the regenerated Thermal Hydraulic loads, (i.e., Deadweight, Seismic, Themal), the results and loads from the as-built analysis were used.

6.2.1 THERMAL HYDRAULIC RESULTS The regeneration of time history thermal hydraulic loads subsequent to the EPRI testing, as previously discussed, was completed. The themal hydraulic analysis was performed based on cold loop seals, the present as-built configuration for both units. Tables 6-1 and 6-2 show the comparisons between the forces previously generated (as-built analysis) and the forces calculated subsequent to the EPRI testing program. These tables illustrate the peak forces encountered by each straight run of pipe during the transient. Table 6-1 compares the forces obtained from the safety valve discharge case. Table 6-2 compares the forces obtained from the relief valve discharge case. The forces included in these two tables are only the forces resulting from the thermal hydrattlic analyses

! and do not account for any other loading conditions or system reaction.

Based on analytical work and tests to date, all acoustic pressures in the upstream piping calculated or observed prior to and during safety valve loop seal discharge are below the maximum permissable pressure.

An evaluation of this inlet piping phenomenum was conducted and the results are documented in a report entitled " Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program", WCAP-10105, dated June 1982. The piping between the pressurizer nozzle and the inlet of the safety valves is 6-inch schedule 160. The calculated maximum upstream pressure for- t'11s size of piping l is below the maximum permissable pressure.

l 6.2.2 STRUCTURAL RESULTS For purposes of providing stress sumaries, the system was broken up into the following three sets of sections:

0440s:10

Section 1: Piping between the pressurizer and the safety valve outlet nozzles (upstream of valves).

Section 2: Piping between the pressurizer and the relief valve outlet nozzles (upstream of valves).

Section 3: Piping between the safety and relief valve outlet nozzles and the pressurizer relief tank (seismically designed downstream portion).

l The stress sumaries for the various loading conditions considered are provided in Tables 6-3 through 6-16. The corresponding node points are shown in Figures 6-1 through 6-8. All stresses listed in the sumary tables are for Unit 2. The stress levels are slightly lower in certain instances for Unit 1. Not tabulating the Unit i results would in no way affect the overall results and conclusions presented in this document.

Our initial evaluation of the new loadings and the old forcing functions subsequent to relief valve discharge indicated that the relief line piping could be qualified upon completion of the structural analysis.

The final structural analyses have confinned and quantified this.

The revised analyses of the PSARV piping subsequent to cold loop seal -

safety valve discharga identified an over-stress region in the piping downstream of the safety valves. This is due to the high magnitude thrust forces immediately upstream of, into and along the common header. The 6-inch safety line branches and the piping imediately upstream of the branch connections are realizing higher than allowable stresses due to these forces. Section 5.2.2.2 of the Joseph M. Farley FSAR was utilized to evaluate this potential overstressed region.

As stated in the FSAR, "A support is provided on the discharge piping as close as possible to each safety and relief valve discharge nozzle so that forces and moments (including pipe whip and reactions following an assumed discharge pipe rupture) will not jeopardize the integrity of the valves, the inlet lines to the valves or the nozzles on the pressur-izer." Based on engineering judgement, the pressurizer itself would not 0440s:10

be affected by the consequences of the overstressed piping if the rupture occurs nor would the operability of the safety valves, inlet lines to the valves or the nozzles on the pressurizer be jeopardired.

6.3

SUMMARY

OF RESULTS AND CONCLUSIONS 4

The evaluation conducted prior to the completion of the structural analysis and based upon the thermal hydraulic loadings subsequent to the simultaneous discharge of all relief valves, the limiting case for relief valve discharge, indicated that the relief line piping could be

qualified upon completion of the analysis for the present as-built l configuration for the J. M. Farley Unit 1 and Unit 2. The structural analyses summarized herein have confirmed and quantified this.

~

The analyses and evaluation of the present system subsequent to the simultaneous discharge of all safety valves identified potential over- ,

stress regions immediately upstream of, into and along the common header near where the safety lines branch in. However, based upon engineering judgement, the analyses results substantiate the fact that the integrity of the valves, the inlet lines to the valves or the pressurizer nozzles is not jeopardized.

t I

0440s

10

TABLE 6-1 UNIT 2 HYDRAULIC FORCE COMPARISONS - SAFETY LINES Table 1 - Safety Line A Table 1 - Safety Line B Force (lbs) Force (lbs)

Location _

Pre Post Location Pre Post i Valve Outlet 9933. 4960. 5 Valve Outlet 23293. 6528.

2 1st to 2nd elbow 24553. 7162. 6 Vertical Run 53699. 37229.

after valve '

7 Run into Header 10344. 39906.

3 Vertical Run 52049. 37205.

4 Run into Header 5907. 36778.

Table 3 - Safety Line C Table 4 - Common Header to PRT I Force (lbs) Force (lbs) location Pre Post Location Pre Post B Valve Outlet 22463. 5956. 14 Header at Safety 15274. 65200.

l Branch Connections 9 Vertical Segment 60526. 29330.

15 Downstream Header 6449. 24279.

10 2nd tt 3rd elbow 6518. 35410.

af ter ,1ve 16 Downstream Header 7683. 22811.

11 Verticai Segment 2518. 38350. 17 Vertical Run to 53730. 51570.

PZR Relief Tank 12 4th to 5th elbow 2702. 39279.

after valve I

13 Run into Header 9423. 35707.

FORCE IDENTIFICATION:

Pre - Force from analysis prior to EPRI program.

Post - Force regenerated in analysis subsequent to EPRI program.

NOTE: ' Location numbers correspond to segment numbers on Figure 4-1.

0440s:10 L

s s.

TABLE 6-2 <

-s UNIT 2 HYDRAULIC FORCE COMPARISON - RELIEF LINE ,

l Pre-EPRI Post-EPRI i Force Location Force (1bs) Force (1bs) 1 Pressurizer nozzle 492. 73.

2 Tee into valve, header 627. 101.

3 6" line north of Tee 514. 63.

4 6 x 3 Reducer 465. 83. \

5 Valve set "A" 6063. 748.

6 Vertical run to common header 11108. 6837.

7 6 inch horizontal run 3630. 5430.

8 Common header /safeky line branches 19165. 3100.

9 Downstream of common header 1385. 1701.

10 Downstream of common header 1414. 1583.

11 Vertical run to 'reifef tank s 8903. 2827.

12 Valve set "B" 2603. 377.

l 13 Vertical 3" section 1684. 871.

14 Tee into 6" vertical pipe 3360. 857. ,

The force location ' numbers correspond to the force numbers on Figure 4-2.

r l

i 0440s:10 ,

o TABLE 6-3 PRIMARY STRESS SLMMARY - UPSTREM OF VALVES Piping System: Pressurizer Relief Line Combination 2 - N + OBE + SOTg Node Maximum All owable Point Piping Component Stress (ksi) Stress (ksi) 1895 Butt weld 27.233 28.98 i 1160 Long radius elbow 28.962 28.98 8

10 90 Tee 19.500 28.98 1890 Reducer 35.85 36.00 5003'. Branch connection 19.873 28.98 1110 Short radius elbow 18.555 28.98 See Tables 2-1 through 2-3 for load combinations and definitions.

\

0440s:10 s s

m TABLE 6-4 PRIMARY STRESS

SUMMARY

- UPSTREAM OF VALVES Piping System: Pressurizer Relief Line Combination 3 - N + SOTg Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1895 Butt weld 24.906 45.0 1150 Long radius elbow 36.278 45.0 1090 Tee 17.980 45.0 1890 Reducer 38.343 45.0 5000 Branch connection 12.827 45.0 l

1110 Short radius elbow 14.333 45.0 See Tables 2-1 through 2-3 for load combinations and definitions.

0440s:10

TABLE 6-5 PRIMARY STRESS

SUMMARY

- UPSTREAM OF VALVES Piping System: Pressurizer Relief Line Combinations 4 and 5 - N + LOCA + SSE + SOTe Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1895 Butt weld 42.595 60.0 1150 Long radius elbow 49.654 60.0 10 90 Tee 24.412 60.0 1890 Reducer 58.318 60.0 5000 Branch connection 23.275 60.0 l

1110 Short radius elbow 20.866 60.0 See Tables 2-1 through 2-3 for load combinations and definitions.

l l

0440s:10

TABLE 6-6 PRIMARY STRESS

SUMMARY

- SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressuri'zer Relief Line Combination 2 - N + SOT g Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1995 Butt weld 20.428 22.56 1975 Long radius elbow 7.502 22.56 i

1340 Reducer 9.147 22.56 l

l 2000 Tee 20.417 22.56 4490 Branch 8.169 22.56 1640 Welded attachment 6.132 22.56 l

' See Tables 2-1 through 2-3 for load combinations and definitions.

l l

i i 0440s:10 1

TABLE 6-7 PRIMARY STRESS

SUMMARY

- SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Relief Line Combination 3 - N + OBE + SOTg Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1995 Butt weld 20.933 29.16 1975 Long radius elbow 9.466 29.16 1340 Reducer 18.G30 29.16 2000 Tee 23.503 29.16 4490 Branch 13.833 29.16 1640 Welded attachment 7.655 29.16 l

See Tables 2-1 through 2-3 for load combinations and definitions.

l 1

0440s:10

TABLE 6-8 PRIMARY STRESS SIMMARY - SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Relief Line Combination 4 - N + SOTg Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1570 Butt weld 60.478* 33.84 1310 Long radius elbow 29.864 33.84 1580 Reducer 90.858* 33.84 2000 Tee 13.480 33.84 4490 Branch 37.498* 33.84 1500 Welded attachment 15.~159 33.84 See Tables 2-1 through 2-3 for load combinations and definitions.

• These points are in the connon header portion.

i l

0440s:10

F

~

9 TABLE 6-9 PRIMARY STRESS

SUMMARY

- SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Relief Line Combinations 5 and 6 - N + LOCA + SSE + SOTe Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 1570 Butt weld 65.119* 45.12 1300 Long radius elbow 40.902 45.12 1570 Reducer 97.674* 45.12 f

l 2000 Tee 19.505 45.12 l

4490 Branch 46.154* 45.12 1500 Welded attachment 23.112 45.12 l

See Tables 2-1 through 2-3 for load combinations and definitions.

• These points are in the common header portion.

l 0440s:10

TABLE 6-10 PRIMARY STRESS

SUMMARY

- UPSTREAM OF VALVES Piping System: Pressurizer Safety Line Combination 2 - N + OBE + SOT g Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi)  !

l l

Butt weld 16.334 36.0 2130 )

2150 Long radius elbow 30.714 36.0 l i 2150 Branch connection 16.506 36.0 )

l 2110 Welded attachment 12.12f, 36.0 See Tables 2-1 through 2-3 for load combinations and definitions.

l 0440s:10

e

.?

TABLE 6-11 PRIMARY STRESS

SUMMARY

- UPSTREAM OF VALVES Piping System: Pressurizer Safety Line Combination 3 - N + SOT g Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 3020 Butt weld 24.779 36.225 3030 Long radius elbow 32.69 36.225 4110 Branch connection 19.428 36.225 4080 Welded attachment 11.019 36.225 See Tables 2-1 through 2-3 for load combinations and definitions.

0440s:10

r-l TABLE 6-12 PRIMARY STRESS

SUMMARY

- UPSTREAM OF VALVES Piping System: Pressurizer Safety Line Combinations 4 and 5 - N + LOCA + SSE + SOTe Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 3020 Butt weld 27.225 48.3 3030 Long radius elbow 46.480 48.3 2150 Branch connection 24.280 48.3 2110 Welded attachment 15.467 48.3 l

l l

See Tables 2-1 through 2-3 for load combinations and definitions.

0440s:10

TABLE 6-13 PRIMARY STRESS SIMMARY - SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Safety Line Combination 2 - N + SOT g Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 4200 Butt weld 5.149 22.56 4420 Long radius elbow 4.655 22.56 4280 Welded attachment 4.260 22.56 4490 Branch 6.988 22.56 See Tables 2-1 through 2-3 for load combinatiers and definitions.

l 0440s:10

I. .

TABLE 6-14 PRIMARY STRESS SIM4ARY - SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Safety Line Combination 3 - N + OBE + SOT'J Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 2240 Butt weld 12.266 33.84 2260 Long radius elbow 12.848 33.84 3350 Welded attachment 7.994 33.84 3480 Branch 14.694 33.84 l

See Tables 2-1 through 2-3 for load combinations and definitions.

0440s:10

r.

TABLE 6-15 PRIMARY STRESS SIMMARY - SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Safety Line Combination 4 - N + SOTg Node Maximum Allowable Point Piping Component Stress (ksi) Stress (ksi) 4390 Butt weld 58.462 33.84 4390 Long radius elbow 72.646 33.84 4360 Welded attachment 75.681 33.84 4490 Branch 95.917 33.84 4370 Clear run 88.954 33.84 l

l i

See Tables 2-1 through 2-3 for load combinations and definitions.

l 0440s:10

T TABLE 6-16 PRIMARY STRESS

SUMMARY

- SEISMICALLY DESIGNED DOWNSTREAM PORTION Piping System: Pressurizer Safety Line Combinations 5 and 6 - N + LOCA + SSE + SOTe Node Maximum Al1owable Point Piping Component Stress (ksi) Stress (ksi) 4390 Butt weld 62.464 45.12 4390 Long radius elbow 77.683 45.12 4360 Welded attachment 78.619 45.12 4490 Branch 113.882 45.12 See Tables 2-1 through 2-3 for load combinations and definitions.

I 0440s:10

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