ML20070B308

TMI-1 Pressurizer Relief Sys - Piping & Support Evaluation Rept

ML20070B308
Person / Time
Site:	Three Mile Island
Issue date:	11/29/1982
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20070B289	List: ML20070B285 ML20070B308
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8212090508
Download: ML20070B308 (50)

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TDR NO. 389 REVISION NO.

PROJECT NO. 412371 PAGE 1

OF -50 I ICONICAL DATA 3lEPORT DEPARTMENT /SECTION Engineering Mechanics Section L ~ TML-1 L RELEASE DATE 10/24-/82 REVISION DATE 10/29/82 DOCUMENT TITLE:

TMI-l Pressurizer Relief System - Piping & Support Evaluation Report ORIGINATOR SIGNATURE DATE DATE Ay g V p ) g ATURE ,

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o DISTRIBUTION ABSTRACT:  !

Brief Statement of Problem J. D. Abramovici NUREG 0737 Item II.D.1 requires an evaluation of safety relief L. R. l..rding and PORV valves. To perform this evaluation GPU participated C. W. Smyth in the EPRI test program to determine the safety relief and E. G. Wallace PORV adequacy under various plant transient conditions. The G. Capodann tests have determined that long inlet conditions lead to J. Colitz valve unstable operation. As a result, GPUN has elected to M. Saxon remove the long inlet (loop seal) and to place the safety A. P. Rochin relief valves on top of pressurizer nozzles. Reanalysis of K. M. Jasani the valves, piping and supports was performed in accordance D. K. Croneberger with EPRI guidance and code requirements. This modification J. Langenbach will be performed prior to restart of TMI-1.

R. Brems (CAI)

R. Toole Summary of Key Results

• R. F. Wilson J. Sheu - The revised piping arrangements were analyzed per code requirements defined in the EPRI test program and stresses

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were found to be acceptable after the addition of one snubber.

- Valve nozzle loads have been reevaluated by the valve vendor and found to be acceptable. ,

- Twenty-eight existing supports had to be reanalyzed of which twenty-five were found to be adequatc "as is" and three require additional redesign. The redesign and implementation will be accomplished prior to restart.

Conclusion s The discharge piping and supports for the safety reliefs and l

PORV are structurally adequate and operational under all j

postulated transient conditions af ter the modifications are performed . .

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t PRESSURIZER RELIEF SYSTEM PIPING AND SUPPORT EVALUATION REPORT FOR THREE MILE ISLAND UNIT 1 GENERAL PUBLIC UTILITIES j

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TABLE OF CONTENTS,,

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l.0 DESCRIPTION OF SAFETY AND RELIEF VALVE INSTALLATION . . . . . . . . . . . . . . . . . . . 1

2.0 DESCRIPTION

S AND RESULTS OF PIPING / SUPPORT EVALUATIONS . . . . . . . . . . . . . . . . . . . . 2 2.1 Discharge Piping Adequacy . . . . . . . . . . . . . . 2 2.2 Conditions Analyzed . . . . . . . . . . . . . . . . . 3 2.3 Safety Valve Backpressure . . . . . . . . . . . . . . 5 2.4 Load Combinations and Acceptance Criteria . . . . . . 5 2.5 SRV Discharge Piping Evaluation . . . . . . . . . . . 5 2.6 PORV Discharge Piping Evaluation . . . . . . . . . . 6

3.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . 7 Appendices Appendix A RELAPS/ MODI Verification Appendix B RELAP5 Models and Discharge Line Piping Forces Appendix C Stress Report m p-4

1.0 DESCRIPTION

OF SAFETY AND RELIEF VALVE INSTALLATION Three Mile Island Unit No. I utilizes a Babcock & Wilcox pressurized water

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reactor (PWR) system with a gross electric power output of 871 MW.

The primary system overpressure protection installation consists of two safety relief valves (SRV) and one power operated relief valve (PORV) as shown in Figure 1 "Three Mile Island Unit No. 1 Relief System Flow Diagram."

The SRV's are spring loaded,self-actuated valves manufactured by Dresser.

These valves are model 31739A. The rated steam relief capacity for each valve is 317,973 lbs/hr.

The original SRV installation included inlet piping with loop seal.

However, the inlet piping and loop seals are being removed and the SRV's are mounted on the pressurizer.

The FORV is manufactured by Dresser. The valve is model 31533VX-30 with 1-3/32" bore. The PORV is mounted directly on the block valve which is mounted on the pressurizer nozzle. The rated steam relief capacity of the valve is 100,000 lbs/hr.

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2.0 DESCRIPTION

S AND RESULTS OF PIPING / SUPPORT EVALUATIONS 2.1 Discharge Piping Adequacy Analysis Procedure:

The integrity of the SRV/PORV discharge piping and their effects on operability of the valves are analyzed for pressure, deadweight, seismic, th.ermal expansion, and valve discharge transient hydrodynamic loadings.

The piping behavior under valve discharge hydrodynamic load is analyzed in three steps:

1. Thermal-fluid analysis to determine the state and flow conditions of the fluid.

. 2. Generation of transient flow forcing functions for piping dynamic analysis.

3. Perform time history piping structural dynamic analysis.

Piping analysis procedures for pressure, deadweight, seismic, and thermal expansion loads are well established and will not be described.

RELAPS/MODL, Cycle 14 (RELAP 5) was used for thermal-fluid analysis.

The RELAP 5 computer code has been verified to be applicable for analysis of the SRV discharge problems by EPRI (Reference 1). The RELNP5 control system is used for generating forcing functions concurrently with the RELAP 5 thermal-fluid analysis execution. These procedures were justified independently by Gilbert Associates, Inc. (GAI) as discussed in Appendix A of this attachment. -

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GAI piping analysis computer code TPIPE version 4.2 was used for the piping analysis. TPIPE is a general piping structural analysis computer codedevelopedbyPMBSystemsEngineering,knc.inSanFrancisco,and has been used by Tennessee Valley Authority and GAI for several nuclear projects.

Piping configuration and support scheme are based on as-built information as modified by Reference 2. Operating conditions analyzed, the load combination method and the results are summarized in the following sections.

2.2 Conditions Analyzed

1. Safety Valve Discharge Transients:

The SRV discharge piping system was analyzed for the following inlet fluid conditions:

a. Saturated steam at 2500 psig,

b. Subcooled water at 4000 F and 2500 psig.

Thermal-fluid analysis was performed for each of the above conditions and the forcing function predictions were evaluated. It was concluded frou the evaluation that the subcooled water discharge case dominates the piping design.

The SRV's on the Three Mile Island Unit No. 1 pressurizer are identical and have exactly the same set-pressure of 2500 psig.

Therefore, it was assumed for the analysis that both SRV's open .

simultaneously and discharge to the 10" tee at the reactor coolant drain tank nozzle:

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(; 7 The inlet pressure transient for the SRV discharge analysis corresponds to the ' feed and bleed' transient which is the limiting subcooled water discharge case as defined by the B&W Plant Condition Justification Report, Reference 3.

Appendix B-1 presents a generalized schematic of the RELAP 5 model used for the hydraulic forcing function generation and the unbalanced piping forces calculated for each case.

The valve inlet conditions and RELAP 5 model initial conditions are given in Table 1.

2. Power Operated Relief Valve Discharge Transients:

The PORV discharge piping system was analyzed for the following inlet fluid condition:

~~~

Subcooled water at 400 F and 2500 psig.

The inlet pressure transient for the PORV discharge piping hydraulic forcing function analysis was determined to be the design worst case based on the EPRI/CE tests which confirmed subcooled water discharges generated higher piping forces than a steam discharge.

The PORV on the Three Mile Island Unit No. 1 pressurizer has a setpoint of 2450 psig.

The inlet pressure transient for the PORV discharge analysis corresponds to the ' feed and bleed' transient which is the limiting subcooled water discharge case as defined by the B&W Plant Condition Justification Report, Reference 3.

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. . 5 g___ Appendix B-2 presents a generalized schematic of the RELAP 5 model used for the hydraulic forcing function generation and the unbalanced piping forces.

The valve inlet conditions and RELAP 5 model initial conditions are given in Table 2.

2.3 Safety Valve Backpressure . .

Backpressure at the SRV's were analyzed for the steady state steam discharge from both SRV's and shown to be:

Valve RC-RVIA - 500 psia.

Valve RC-RVIB - 490 psia.

2.4 Load Combinations and Acceptance Criteria

___ Load combinations and acceptance. criteria for the safety and relief valve piping evaluation of Three Mile Island Unit No. I were based on " Load Combinations and Acceptance Criteria for the Safety and Relief Valve Piping Evaluation," Reference 4.

2.5 SRV Discharge Piping Evaluation

.The discharge piping and supports have been evaluated with the following conclusions and recommendations:

1.= The maximum stress will not exceed the allowable limits af ter adding one 3 Kip snubber at joint 346 in the Z direction.

b.

1 SRV (ME-88. 89) - Maximum stress in this piping branch has just reached but is still within code allowable limits (See Table 6.2).

Also, with the design modifications performed on existing supports PR-28 & PR-33, all supports will be within their original design capacities.

FCC c. SRV (ME-91, 92) - Maximum stress in this piping branch is within

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code allowable limits (See Table 6.3). Also, with the design mcdifications performed on existing support PR-32 all supports will be within their original design capacities.

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2. Tha valva nozzla lecda et tha cutlat fitngis, imp'ossd by tha discharge piping system, exceed the allowable shown in the vendor catalog (Reference 6). These loads have been re-evaluated by Dresser and found acceptable.

3. The reactor coolant drain tank (k'DL-T-3) nozzle load is within the allowable limits, calculated by GAI.

For details of the piping and support evaluation, refer to the Piping Stress Report (Appendix C).

2.6 PORV Discharge Piping Evaluation The discharge piping and supports have been evaluated with the following conclusions and recommendations:

,.. 1. The valve nozzle load at the outlet flange imposed by the discharge piping system, exceeds those shown in the previous design. The valve load has been re-evaluated by Dresser, and found acceptable.

2. PORV (ME-162, 93) - Maximum stress in this piping branch is within code allowable limits (See Table 6.1) . Also, a ll existing supperts are within their original designed capacities.

For details of the piping and supports evaluation, refer to the Piping Stress Report (Appendix C).

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3.0 REFERENCES

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1. Application of RELAP5/ MODI for Calculation of Safety and Relief Valve Discharge Piping Hydraulic Loads",# EPRI, April 1982.

2. GPU Pressurizer Relief Rip-Out and Modification Study, Drawings ID-220-03-001, -002' (preliminary) .

3. Valve Inlet Fluid Conditions for . Pressurizer Safety and Relief Valves for B&W 177- and 205-FA Plants, April 1982.

4. Guide for Application of Valve Test Pr.ogram Results, EPRI, April 1982.

5. Safety and Relief Valve Test Report, EPRI, April 1982.

6. Dresser, N100 Pressurizer Safety and Relief Valve Catalog, pg. 15.

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s g-- TABLE 1 SRV ANALYSIS Valve Inlet Conditions RELAP5 Model Initial Conditions Safety Valve Rated Capacity Dresser 31739 A 317,973 lbs/hr Valve Inlet Conditions: Case 1 Case 2 Pressurizer Pressure 2500 psig 2500 psig Pressurizer Temperature Saturated 4000F (subcooled)

Valve Opening Time (Reference 5) 12 msec 40 msec Stroke vs. Time Linear Linear RELAPS Model Initial Conditions:*

Discharge Line Pressure 55 psig 55 psig Reactor Coolant Drain Tank Pressure 55 psig 55 psig Discharge Line Temperature Saturated Saturated

• Note: Discharge line initial conditions based on 55 psig (rupture disc burst pressure) yielded higher piping loads than initial conditions of 0 psig, 120 F, and air in the discharge lines.

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. l TABLE 2 L-PORV ANALYSIS Valve Inlet Conditiohs RELAP5 Model Inie'.ai conditions Dresser 31533VX-30 Relief Valve Rated Capacity 100,000 lbs/hr Valve Inlet Conditions:

Pressurizer Pressure 2500 psig Pressurizer Temperatbre 4000F (subcooled)

Valve Opening Time (Reference 5) 60 maec Stroke vs. Time Quick Opening RELAP5 Model Initial Conditions:

Discharge Line Pressure 55 psig Reactor Coolant Drain Tank Pressure 55 psig Discharge Line Temperature Saturated

• Note: Discharge line initial conditions based on saturated steam

@ 55 psig (rupture disc burst pressure) yielded higher piping loads than initial conditions of 0 psig, 120 F, and air in the discharge lines.

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Appendix A

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RELAP5/H0Dl: Verification with EPRI/CE test program data. Application to saf tey/ relief valves discharge transients analysis.

RELAPS/ MOD 1 (Reference 1) is an advanced one dimensional thermal-hydraulic system analysis code developed for the U. S. Nuclear Regulatory Commission by EG&G Idaho, Inc. The code differs from the well known RELAP4 code in that it is based on a non-homogeneous, aon-equilibrium "

hydrodynamic model as opposed to the homogeneous equilibrium model used by RELAP4.

While RELAPS/ MOD 1 predicts the time dependent thermal-fluid conditions in the safety / relief valve discharge piping for a given postulated transient, the hydrodynamic pipe forces induced by the transient flow CIT are not directly computed. Forcing functions were generated for each bounded straight pipe segment in the discharge piping system solving the acceleration term of the momentum balance equation. These net or wave force calculations can be performed within RELAPS/ MOD 1 by using the program's control system which provides the user with the capability to evaluate algebraic and ordinary differential equations. Prior to the availability of EPRI/CE PWR Saftey Relief Test Program data, this methodology was verified at Gilbert Associates by analyzing a steam blowdown transient and comparing the results to those published in the literature (Reference 2 and 3). The RELAP5/ MOD 1 results are in excellent-agreement with the published data. The adequacy of RELAP5/ MODI for calculation of safety / relief valve discharge piping hydrodynamic loads has been demonstrated by Intermountain Technologies, Inc. (ITI) and T(( documented in (Reference 4). The ITI effort included modeling of the EPRI/CE O

safety / relief ' valve test stand (Figure 1) with RELAPS/UODI and' comparison of results with the test data. The ITI results indicate that RELAP5/ MODI can provide good engineering estimates of pipe loads. induced by safety / relief valve discharge transients.

As an additional verification of the forcing function generation and modeling techniques used in the analysis of the Three Mile Island Unit # 1 safety /

relief valve discharge piping, the EPRI/CE test no. 917 (See Reference 5) was simulated with RELAPS/ MODI and the results compared to the test results. '

Figures 2-4 show a comparison of two test pressures (PT08 and PT10,' See Figure 1 for location) and one test pipe load (F2 , see Figure 1 for location) to RELAP5/ MODI results. A direct comparison of other pipe segments loads to RELAPS/ MODI results was not possible because the segments were not rigidly supported and thus require a piping structural analysis.

i The comparison on Figures 2 & 3 indicates that RELAPS/ MODI can adequately simulate the transient thermal-fluid conditions in the discharge piping following safety / relief valve opening. Figure 4 shows the good agreement between the measured load on the vertical discharge pipe segment and the one predicted by RELAP5/ MODI and the forcing function methodology described earlier. The same techniques that were used in this verification were then applied in the RELAPS/ MODI modeling of the Three Mile Island Unit #1 plant pressurizer safety / relief valve discharge system and determination of pipe forcing functions.

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REFERENCES C

1. V. H. Ransom, et.al., "RELAP5/ MODI Code Manual," Vols. 1 and 2, EG&G Idaho, NUREG/CR-1826, March 1981.

2. B. R. Strong, Jr., and R. J. Easchiere, " Pipe Rupture and Steam / Water Hammer Design Loads for Dynamic Analysis of Piping Systeus," Nuclear Engineering and Design, Vol. 45, 1978, p.419 - 428.

3. V. R. Burke,and S. W. Webb, "RELAP4/ THRUST Computer Code Manual."

Gilbert Associates, Inc., March, 1980.

4. " Application of RELAP5/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydraulic Loads", EPRI, April, 1982.

5. " Safety and Relief Valve Test Report", EPRI, April 1981.

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RELAP5 MODELS AND DISCHARGE LINE PIPING FORCES Safety Valve Discharge Transients, RELAP 5 Models and Results The safety valve discharge lines were modeled using a 233 volume and 233 junction RELAPS model.

Since the valve setpoints are both 2500 psig, the hydraulic forces on the discharge lines were evaluated assuming both safety valves open simultaneously. The generalized RELAPS model is given in Figure B-1.1.

The hydraulic forcing functions are defined in Figure B-1.2 with the positive force direction defined opposite to the direction of flow.

Figures B-1.3 presenu the unbalanced forces along a straight segment of pipe for the 4000F subcooled water discharge

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transient.

Power Operated Relief Valve Discharge Transients, RELAP5 Model and Results The PORV discharge line RELAPS model contains 147 volumes and 146 junctions.

The generalized RELAP5 model is given in Figure B-2.1.

The hydraulic forcing functions are defined in Figure B-2.2 with the positive force direction defined opposite to the direction of flow.

Figures B-2.3 presents the unbalanced forces along a straight segment of pipe for the 4000F subcooled water discharge transient.

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RELAPS MODELS AND DISCHARGE LINE PIPING FORCES ,,

Safety Valve Discharge Transients, RELAP 5 Models and Results The safety valve discharge lines were modeled using a 233 volume and 233 junction RELAPS model. Since the valve setpoints are both 2500 psig, the hydraulic forces on the discharge lines were evaluated assuming both safety valves open simultaneously. The generalized RELAPS model is given in Figure B-1.1.

The hydraulic forcing functions are defined in Figure B-1.2 with the positive force direction defined opposite to the direction of flow.

Figures B-1.3 presents the unbalanced forces along a straight segment of pipe for the 4000F subcooled water discharge transient.

Power Operated Relief Valve Discharge Transients, RELAPS Model and Results The PORV discharge line RELAPS model contains 147 volumes and 146 junctions. The generalized RELAPS model is given in Figure B-2.1.

The hydraulic forcing functions are defined in Figure B-2.2 with the positive force direction defined opposite to the direction of flow.

Figures B-2.3 presents the unbalanced forces along a straight segment of pipe for the 4000F subcooled water discharge transient.

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ABSTRACT l

The purpose of this report is to present the result of i

piping structural analysis of pressurizer safety / relief valve discharge line due to fluid transient, seismic, thermal and deadweight effects, also to demonstrate the evaluation of piping system in accordance with the guideline issued by EPRI, April 1982 (" Guide for application of valve test program results to plant specific evaluation", Appendix E.)

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w TABLE OF CONTENTS Section Page

1. System Description - - - - - - - - -

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2. Analysis Description --- - -- -

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3. Analysis Tool - - 4

4. Transient Conditions Analyzed - -

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5. Loading Combination and Acceptance Criteria ----- 6

6. Sunanary of Results and Conclusions -- ----- 8

a. piping stress

b. pipe support load

c. nozzle load

d. valve acceleration t_ 7. Design Input - -- -

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8. References -

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1. SYSTEM DESCRIPTION Three Mile Island Nuclear Station, Unit-1 utilizes a Babcock & Wilcox prescurized water reactor system with an electric power output of 871 MW.

Its primary loop overpressure protection system consists of two safety relief valves (SRV) and one power operated relief valve (PORV). Both SRVs and PORV are installed directly on the pressurizer, and consequently the valve connections at the pressurizer are assumed as anchors.

The two 6", SRV discharge lines and the one 4" PORV

___ discharge piping merge to a 10" common header, 1.5' above the reactor coolant drain tank. -

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2. ANALYSIS DESCRIPTION The integrity of the piping system and its effects on the operability of the valves are analyzed for pressure, dead-weight, seismic, thermal expansion, and the flow transient loads. The effects of the flow transient loads are analyzed in three steps: (1) perform thermal-fluid analysis to determine the states and the flow conditions of the fluid, (2) generate time dependent, flow induced forces on the pipe for piping dynamic analysis, (3) perform piping struc-tural dynamic analysis. Steps (1) and (2) are performed by

AEA - Dept. of Gilbert Associates, Inc. and are reported separately. This report covers step (3) and the analysis for pressure, deadweight, seismic and thermal expansion loadings. Fig. 2.1 shows the major analysis procedure.

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TPIPE PSLAP-5 ANALYSIS ANALYSIS

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(therzal-structural) hydraulic) on CDC on UCC

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, computer system (performed by set up vipinJt geometry model AEA Dept) 9 C

temperature, and forcing function deadveight analysis a

thermal expansion analysis II transfer to TPIPE format in U IBM system, GAI seismic analvsis i U V rime historv analvsis  : trassr.it to CD;' computer service y

.n TPIPE post process (piping stress and supports evaluation

.; per EPRI report " Guide for application of valve test program results to plant specific evaluation ", Appendix E.)

1 u p p u pipe support equipment valve stress load nozzle acceleration evaluation evaluation evaluation evaluation i

l conclusions and recommendations Fig. 2.1 Analysis Functional Flow Diagram p oem

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3. Anal'ysis Tool Gilbert / Commonwealth piping analysis computer code TPIPE version 4.2, July 1982, was used in the piping system analysis.

TPIPE is a general piping analysis computer code developed by PMB Systems Engineering, Inc. in San Francisco, and has been used by Tennessee Valley Authority, Gilbert /Consnonwealth and other organizations for several nuclear power plant projects.

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4. Transient Conditions Analyzed The SRV discharge piping could be subjected to saturated discharge and subcooled water discharge. Both cases were analyzed separately and the worst of the two loading cases for piping stresses and support loads were screened using TPIPE postprocessor.

The PORV discharge piping was analyzed for a 400F, 2500 psig subcooled water discharge.

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5. Loading Combination and Acc1ptance Criteria

. The loading combination and acceptance criteria for the SRV and PORV discharge systems were based on table 5.1, which were given in appendix E of reference 6.

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LOAD COMDINATIONS AND ACCEPTANCE CRITERIA FOR PRESSURIZER SAFETY .

AND RELIEF VALVE PIPING AND SUPPORTS - SEISMICALLY. DESIGNED DOWNSTREAM PORTION (TABLE 5.0 Plant / System Service Stress Combination Operating Condition Load Combination Limit Normal N A 1

Upset N + SOTg B 2

Upset N + OBE + SOT g C 3

4 Emergency N + SOT E Faulted N + MS/FWPB or DBPB D 5

+ SSE + SOT p Faulted N + LOCA + SSE.+ SOTp D 6

I DEFINITIONS OF LOAD ABBREVIATIONS N = Sustained Loads During Normal Plant Operation SSE = Safe Shutdown Earthquake

• l SOT = System Operating Transient MS/FWPB = Main Steam'or Feedwater Pipe. Break SOT U

= Relief Valve Discharge Transient [1) DBPB = Design Basis Pipe Break SOT E = Safety Valve Discharge Transient y LOCA = Loss of Coolant Accident SOT p = Max (SOTU; SOTE); or Transition Flow OBE = Operating Basis Earthquake This table s applicable to the seismic. ally designed portion of downstream non-Category I piping (and supports) necessary to isolate the Category I portion from the non-seismically designed piping response, and to assure acceptable valve loading on the discharge nozzle.

Use SRSS for combining dynamic load responses.

6. Summary of Results and Conclusions 6.1 The results of the analysis of the PORV/SRV discharge piping are summarized as follows:

a. PORV (ME-162, 93) - Maximum stress in this piping branch is within code allowable limits (See Table 6.1) . Also, all existing supports are within their original designed capacities.

b.

SRV (ME-88, 29) - Maximum stress in this piping branch is ,

summarized on page 10 (See Table 6.2). Also, with the design modifications performed on existing supports PR-28 & PR-33, all supports will be within their original design capacities.

c. SRV (ME-91, 92) - Maximum stress in this piping branch is within code allowable limits (See Table 6.3). Also, with the design modifications performed on existing support PR-32, all supports will be within their original design capacities.

d. The SRV nozzle loads at outlet flange imposed by the piping were evaluated by Dresser and found acceptable.

c. The PORV nozzle was evaluated by Dresser and found acceptable.

f. The Reactor Coolant Drain Tank (WDL-T-3) nozzle loads are below the allowable limits used for previous analysis,

g. The normal operating condition for the S/RV discharging piping is when the SRV and PORV are closed. The discharging piping temperature will be the same as the plant normal environmental temperature. The discharging piping will not contain any sub-cooled water during the system normal operating condition.

t Table 6.1 - Stress Summary Of PORV Branch Piping (ME-162,93) '

4 TP PE Tnt put seau otscwARCE P!P Nc e mC-162.931 sw.0..e4se64eee t LLCwou FRon PORutRC-Rv21 10 PRE 55uRIZER RELIEF TAmt ASRE C0DE CLASS 2 57RCSS SUMMARv

+

nEngER n0DAL COM CODC ALL0 Waste STRESS DESCRIP710M NanE MARE Mo. STRESS- STRESS RATIO 1540 3 146 s B 6159. 15500. 4e MAE STRESS 1500

• 146 8 5 6150. 15500. 40 MAN STRESS RATIO 1274 8 CENTAS to 24722. 270 .92 RAX STRESS 1270 CENTAS le 24722. 27.00.

60. .92 RAM $ TRESS RATIO 1274 I CENTat 11 27816. 4250s. .65 MAX STRESS 1270 3 CENTRI 11 27816. 42500 .65 MAX STRESS RATIO 1510 3 148 5 9U 16762. 19609. .90 MAX STRESS 1518 3 148 8 9U 16762. 19608. .90 RAM $ TRESS RATIO 1510 a les 3 9E 21122. 27904. 76 RAX STRESS 1588 t 14e 2 9E 21122. 76 MAX STRESS E1710 1518 3 148 I 9F 29785. 7'.'9 k 20s. 8 4. .50 MAX STRESS 1510 s 148

• 9F 29785. 37200. .Se RAM STRESS RATIO 1270 CENTRE PR 23678. 36480. .79 MAM STRESS 1870 8 CENTRE PR 23678. 36480. 79 max STRESS RATIO 15 e les s AU 41608. 680000. .07 max STRESS 1510 8 148 !s AV 4160s. 6teese. .47 MAN STRESS RATIO TOTAL NUMSER OF P!PC REMBER$ WITH NODAL POINT $ CREATER THAM A THRESM0LD STRESS RATIO 0F t.004 ,

EQUATION 3..... O EQUATION 19..... 9 EQUATION 11..... 4 ECUATION 9U..... 8 EQUATION 9E..... 0 EQUAT10N 9F..... O P!PE RUPTURE.... 8 ACTIVE WALVE.... e em cy 0

F V

--.-. ... ..e -. .. . .

Table 6.2 Stress Sunsnary Of-SRV Branch Piping (ME-88,89)

, . . . . . . . . . . . . . . . . - . - - - -- .~ . * ==~ ~

TP!PC TM! PUR SERV DISCHAeGC P!P!NC (MC= 88.09) 10.0.*e450640005 LLCMOU

.. FROR $RVtMC=147) asnC eooC TOeLAss PRES $URIZER 2 RC[s!CF vRCss TAMC suMMARv

. . MEMBER MODAL COM CODE ALLOUASLC STRESS DESCRIPTION MARC MAMC MO. STRESS STRCSS RATIO 2250 3 223 3 3 6661. 15500. .39 MAX STRESS 2250 3 223 K S 6061. 15584. .39 MAX STRESS RATIO 2210 3 219 3 to 16087 27ete. .se Max STRESS 221e 3 219 3 10 . 16987. 27000. .60 MAX STRESS NATIO 2250 2 CENTAS 11 21549. 42580. .51 MAX STRESS 225e 8 CENTRt 11 21549. 42500. .51 MAX STRESS RATIO i 2410 2 240 8 9U 19574. 18604. .55 MAN STRESS 2410 3 240 3 90 19874 18644. .55 MAX STRESS RATIO 247e 3 245

• 9E 20370. 27900. 73 RAW STRESS 2474 3 245 3 9E 20370 27900. .73 MAX STRESS RATIO

, 2470 3 245 3 9F 37275. 37200. 1.80 MAM STRES$

i 247e 3 245 3 9F 37275. 37200. 1.08 Max STRESS RATIO 2250 3 CENTRE PR 24390. 36480. .57 MAN STRESS 2250 s CEMTRE PR 24390. 36480. .G7 MAX STRESS RATIO 2470 3 245 s AU 35577 Seette. 46 MAX STRESS 2474 3 245

• AV 38877. 680000. .06 MAX STRESS RATIO TOTAL NUMBER OF PIPC MEMBER $ UITH MODAL POINTS CREAiit THAM A THRESHOLD STRESS AAT!0 CF 1 800 COUATION 5..... e EQUATION 10..... O EQUATICM 11..... 4 l EQUATICM 90..... 4 EQUA?!0M 9C..... O COUATION 9F..... 1 PIPE RUPTURC.... e ACTIVE WALUC.... 0 .

l .

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"-~~Tslil'e T.'3~~~~~'StfesTSunimiiry o~f*'SRV'Tiaiich P~ii i~n'g i (ME-91,~9'27- j i--

~~TPIPE TMI PWR S/RV DISCHARGE PIPING (ME- 91,92) (W.D.2045064000)' LLCHOU

....FROM SRV(ME-148) TO PRESSURIZER RELIEF TAHK ,

_. LLC TMI/EPRI PWR/PORV PIPING EVALUATION (ME91 ,ME92)

'i

__A S M E C0DE CLA55 2 $ TRESS

SUMMARY

MEMBER HDDAL EQH CODE' ALLOWABLE STRESS DESCRIPTION i HAME NAME HD. STRESS STRESS .

RATIO l

~~

3150 x 314 x 8 5625. 15500. .36 MAX STRESS

. 3150 x 314 W 8 5625. 15500. .36 MAX STRESS RATIO 3290 x 328 N 10 26258. 27000. .97 MAX STRESS l

. 3290 M 328 N 10 26258. 27000. .97 MAX STRESS RATIO i

~

3290 M 328 x 11 30145. 42500. .71 MAX STRESS

.- 3290 x 328 x 11 30145. 42500. .71 MAX STRESS RATIO 3190 x 318 W 9U 8718. 18600. .47 MAX STRESS 3190 x 318 N 90 8718. 18600. 47 MAX STRESS RATIO 3070 x 306 x 9E 19900. 27900. .71 MAX STRESS j, 3070 W 306 x 9E- 19900. 27900. .71 MAX STRESS RATIO  :

3070 x 306 x 9F 35129.- 37200. . .94 . MAX STRESS

.-. 3070 M 306 M 9F 35129. 37200. .94 MAX STRESS RATIO -

3290 x 328 N PR 32868. 36480. .90 MAX STRESS

,--- 3290 N 328 M PR 32868. 36480. .90 MAX STRESS RATIO 3310 M 329 M AV 24597. 600000. .04 MAX STRESS

--- 3310 M 329 M AV 24597. 600000. .04 MAX STRESS RATIO

~'

TOTAL HUMBER OF PIPE MEMBERS WITH NODAL POINTS GREATER THAH A THRESHOLD STRESS RATIO EQUATIDH 8..... 0 EQUATIDH 10..... 0 EQUATION 11..... 0 EQUATION 9U..... 0 EQUATION 9E..... 0 EQUATION 9F..... 0 .

PIPE RUPTURE.... 0 . . .

ACTIVE V ALVE. . . . 0

. .. --...- _ . _ . . _ . _ 2.u.__.

ne*

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7. DESIGN INPUT 7.1 Assumption

a. Piping ' filled with water was used in the completed analysis (DW + THERMAL + OBE + BLOWDOWN). Additionally, piping filled with steam was analyzed in deadweight case only.

b. Three piping branches are structurally independent of each other.

c. Valves were simulated as the same. size pipe but doubled pipe thickness.

d. 450 F was used to piping thermal expansion analysis. (Ref.

to AEA's RELAP 5 analysis.)

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5. REFERENCE .

1. Dresser Valve N-100 Cataloge, " Consolidated and Hancock Valves for Nuclear Service".

2. Safety and Relief Valve Test Report, EPRI, April 1, 1981.

, 3. TPIPE 4.2 User Manual, June 1982.

4. Seismic Design Criteria, October 9, 1969.

f 5. Valve Selection / Justification Report, April 1, 1982.

6. " Guide for Application of Valve Test Program Results to t Plant Specific Evaluation", Rev.1. April 5,1982.

l f-.-.-

- l8-1