ML20203J306

Reactor Vessel Thermal Cycle Fatigue Assessment

ML20203J306
Person / Time
Site:	Peach Bottom
Issue date:	01/20/1986
From:	Branlund B, Ranganath S, Stevens G GENERAL ELECTRIC CO.
To:
References
MDE-166-0785, MDE-166-785, SASR-85-54, NUDOCS 8608050235
Download: ML20203J306 (26)

Text

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DRF 137-0010

, SASR 85-54 REV.1 H ,

MDE-166-0785 I

l 3 .; J - / 7 7 9f n 76 O PEACH BOTTOM UNITS 2 AND 3 REECTOR VESSEL THERMAL CYCLE FATIGUE ASSESSMENT O

Prepared by: O 3.1%

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B. J. Branlund, Engineer Structural Analysis Services Verified by: kI $4 G. L.U Stevens, Engineer l/Zo/SC, Structural Analysis Services O Approved by: N^ ^ ^ ^--

S. Ranganath", Manager Structural Analysis Services Approved by: [ de O A. E. Rogers, Manager Application Engineering Services O

GENERAL h ELECTRIC 8608050235 860120 0 PDR ADOCK 05000277 O P PDR

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,(} . IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please read carefully The only undertakings of General Electric Company respecting information in this document are contained in the contract between

() Philadelphia Electric Company and General Electric Company, as identified in the purchase order for this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General El'ctric e Company makes no

() representation or warranty, and assumes no-liability as to the completeness, accuracy, or usefulness of the information contained in this document.

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O CONTENTS

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1.0 ABSTRACT 1-1 (3

2.0 INTRODUCTION

2-1 3.0 COMPONENT SELECTION 3-1 4.0 ALLOWABLE CYCLE CALCULATION 4-1

() 5.0 RESULTS AND CONCLUSIONS 5-1

6.0 REFERENCES

6-1 APPENDIX A ADDITIONAL CYCLES FOR THE FEEDWATER A-1 N0ZZLE O ILLUSTRATIONS FIGURE 1 SAFETY RELIEF VALVE BLOWDOWN EVENT 2-3 FIGURE 2 FEACTOR PRESSURE VESSEL THERMAL CYCLES 3-5 O FIGURE 3 RECIRCULATION INLET AND OUTLET N0ZZLE 3-6 THERMAL CYCLES ,

FIGURE 4 STEAM OUTLET N0ZZLE THERMAL CYCLES 3-6

,(3 FIGURE 5 FEEDWATER N0ZZLE THERMAL CYCLES 3-7 FIGURE 6 CRD HYDRAULIC SYSTEM RETURN N0ZZLE 3-8 THERMAL CYCLES l

l FIGURE 7 CONTROL ROD DRIVE N0ZZLE THERMAL CYCLES 3-8

• TABLES TABLE 1 MAXIMUM FATIGUE USAGE FACTORS 3-10 TABLE 2 ALLOWABLE NUMBER OF ADDITIONAL SRV EVENTS 4 O O

! 11 O

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1.0 ABSTRACT 4> ..

This report evaluates the effect of additional SRV blowdown cycles on the reactor pressure vessel components. The evaluation involved two

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tasks:

3

1. Determination of components that are affected the most by the additional cycles.

4

2. Determination of the allowable number of cycles for these Components.

l The evaluation shows that the feedwater nozzle is affected the most by the additional cycles. Based on the limiting results a total of Q

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114 additional SRV blowdown cycles can be allowed during the 40 year

! plant life without exceeding Code fatigue usage limits.

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2.0 INTRODUCTION

() .

Design of nuclear power plant comp,nents is based on providing adequate margins on rupture and fatigue initiation. The fatigue margin is maintained by assuring that the fatigue usage, based on an assumed cyclic duty and a fatigue design curve, is less than one. The number of

)

cycles and the magnitude of the pressure and temperature transients are described in the design thermal cycle document. The fatigue usage calculations are based on the design duty document (Reference 1).

Although the cycles described in this document are based on BWR plant experience, they are at best estimates and are intended only as the design basis.

s Sometimes the actual number of cycles for a given transient may J

exceed the number of cycles specified in the design basis. In such cases it is necessary to consider the projecteel number of cycles for the total design life based on current experience and demonstrate that the fatigue usage limits are satisfied.

The safety relief valve (SRV) blowdown transient at Peach Bottom is -

an example of an event where the projected number of cycles are likely to exceed the number specified in the design basis. Thus, it is O necessary to demonstrate that the fatigue usage for the projected number of cycles including the additional SRV blowdown cycles is lees than one.

An alternate way of demonstrating adequacy is to determine the maximum number of SRV blowdown cycles that can be tolerated without exceeding

.O the fatigue usage limit of one. This is often possible since the fatigue usage for the original design cycles is often much less than one.

O The original design basis explicitly considered two SRV blowdown cycles with total depressurization (References 1 and 2). In addition to this there are a total of 198 other scrams (including loss of feedwater pumps, turbine generator trip and 147 other scrams). Although the number of Peach Bottom SRV blowdown cycles are expected to be greater than the number considered in the design basis, many of the SRV events 2-1 l

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,, will probably involve a smaller pressure drop and would not have a

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startup following the event. However, to determine the allowable number of additional SRV blowdown cycles, we will assume that the event involves depressurization to 50 psig as shown in Figure 1 followed by a startup and turbine roll. This is in addition to the other scrams.

O startup, and turbine roll events already considered in the original analysis.

The assumption of startup/ turbine ro'll following depressurization is important since the highest stress range influencing the transient comes from the turbine roll rather than the SRV blowdown. Thus, differences in the SRV blowdown event (i.e. rate of cooling during the blowdown) are not significant from the fatigue viewpoint. For example, c) if the cooldown rate from 375'F to 100*F in Figure 1 were 125*F/hr instead of the assumed 100*F/hr the overall fatigue usage would not be significantly affected.

O This report will evaluate the effect of these additional SRV blowdown cycles on the reactor pressure vessel components. This evaluation involves two tasks:

O 1. Review the thermal cycle diagrams and original design analyses of the reactor pressure vessel components and determine the components that will be affected the most by the additional cycles.

v

2. Determine the allowable number of additional cycles for the most limiting component. This should be. based on the fatigue usage in the original Code design stress report (Reference 3).

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3.0 COMPONENT SELECTION O

The first task is to review the thermal cycle diagrams and original design analysis for each component and determine the components that will be affected the most. The following two criteria were used to 8 elect the c 2Ponents that are significantly effected by the additional

O SRV blowdown events.

1. Components that accumulate a large incremental fatigue usage g for the additional cycles. These are typically components with a significant thermal transient. This will be apparent from examination of the thermal cycle diagram.

g 2. Components with a 40 year fatigue usage greater than 0.30 before considering the additional cycles. Any component with a fatigue usage less than 0.30 has a margin of three times in addition to the safety factor of three inherent in the fatigue

""*1 7 ****

O 3.1 THERMAL CYCLE EVALUATION Corponents that accumulate a large incremental fatigue usage due to the additional cycles were identified by examining the thermal cycle diagram for each component and evaluating the impact of the transient.

The evaluation was based on the expected stress range.

O 3.1.1 Reactor Pressure Vessel The additional cycles for Regions A, B, and C are shown in Figure 2. The SRV blowdown and startup events cause the pressure O

stresses to change from normal to zero, but are not large enough to to significantly affect the fatigue usage.

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O 3.1.2 Recirculation Inlet and Outlet Nozzles C) .

The transient definitions for these two nozzles are the same for the startup, turbine roll and SRV blowdown events. The flow is affected by these transients ao shown in Figure 3, while the temperature and pressure is the same as Region B. Therefore, as discussed in the Reactor Vessel Section the fatigue usage is unaffected by the additional transients.

, 3.1.3 Steam Outlet Nozzle The temperature and pressure for this nozzle is the same as Region A. The flow shown in Figure 4 is steam flow. Since the temperature and pressure are the same as Region A the fatigue usage will be unaffected as discussed in the Reactor Vessel Section.

3.1.4 Feedwater Nozzle Thermal Cycles

.A The startup and SRV blowdown events are similar to or less severe than those shown for Region A. However, as can be seen in Figure 5, the turbine roll event is clearly significant. As discussed earlier the cooldown rate during the blowdown does not influence the fatigue O results, since the turbine roll stresses are controlling.

For the turbine roll event it is conservatively assumed that the water in the feedwater line is 50*F. No credit is taken for feedwater heating O

following the SRV blowdown. Therefore, during a turbine roll the water in the feedwater nozzle undergoes a step change from 552*F to 50*F.

This step change in temperature will cause high stresses to occur in the nozzle due to the thermal shock. This could significantly affect the O fatigue usage. It should be noted that the Feedwater Nozzle and Safe End were analyzed using a design bcsis with a larger number of thermal cycles (Reference 3).

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O 3.1.5 Drain, Core Spray, and Head Spray Nortles O .

The temperature and pressure for the drain, core spray and head spray nozzles are respectively the same as Region C, Region B, and Region A. The drain flow is used to drain cold water out of Region C to n bring the warmer Region B water down into Region C so that the v

temperature difference between these two regions will not exceed the Reactor technical specification limits. There is no flow in the core spray or head spray nozzles during the startup, turbine roll, or SRV blowdown events. Therefore, as discussed for the Reactor Vessel Section the additional transients will not affect the fatigue usage.

3.1.6 CRD Hydraulic System Return Nozzle O

When the control rods are withdrawn to start reactor warmup, nozzle flow rises to 1.2 gpm and the temperature drops from 70'F to 50'F, the temperature of the condensate storage tank. Since the nozzle remains at 50*F through the turbine roll, the effect of these two transients on fatigue usage will be insignificant. During an SRV blowdown event, however, after a reduction of temperature to 50*F, the flow is turned -

off for sufficient time that the nozzle comes to thermal equilibrium with Region B. Therefore, the temperature will cycle between 546*F and n~

50*F as shown in Figure 6. This transient is sufficient to increase the fatigue usage for the additional events. The CRD return nozzle has been capped in the Peach Bottom Units, therefore, the actual fatigue cycling on this nozzle may be significantly lower.

l l 3.1.7 Instrumentation and Core Differential Pressure and Liquid l Control Nozzles O

There is generally no flow in these nozzles during the startup, turbine roll, and SRV blowdown events. Therefore, the fluid temperatures and pressures correspond to the vessel region temperatures and pressures to which the nozzles are attached. Consequently, the fatig.se usage is unaf fected.

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O 3.1.8 Control Rod Drive Nozzle-During the startup and turbine roll events the flow is constant at

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0.34 gpm with a temperature of 60*F. Consequently, the f atigue usage is unaffected. However, during the SRV blowdown event, the flow of cold O vater into the vessel nozzle stops and water flow reverses for 2.3 seconds (see Figure 7). The fluid temperature undergoes a step change from 60*F to that of the Region C water. This temperature may be as high as 546*F, the maximum temperature of' Region C. When CRD flow into A'

the vessel is re-established, the temperature is assumed to be 50*F rather than 60*F because of the higher flow rate. These latter conditions are assumed to prevail long enough for thermal equilibrium to be established and then the flow drops to 0.34 gpm and the temperature O increases to 60*F. This transient is sufficient to increase the fatigue usage for the additional cycles.

3.1.9 Conclusion v

Based on the thermal transient evaluation for each component, the -

feedwater, CRD hydraulic return, and CRD nozzles should be evaluated.

Since components could have high fatigue for reasons other than thermal O transients, selection based on the second criterion, i.e. available fatigue margin, may increase le component selection.

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FIGURE 3 RECIRCULATION INLET AND OUTLET N0ZZLE THERMAL CYCLES O

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l 3.2 FATIGUE USAGE EVALUATION l

C) ..

Fatigue usage was also used as a criterion for component selection.  ;

Those components with a 40 year fatigue usage greater than 0.30, before l considering the additional cycles, were evaluated. Maximum fatigue usage factors for each of these components are shown in Table 1. Using

)

the second criterion the number of components to be evaluated can be increased from three to five. The five components are:

Component Original Design Analysis

)

Fatigue Usage Feedwater Nozzle 0.89 CRD Hydraulic Return Nozzle 0.36 Support Skirt 0.55

)

Refueling Containment Skirt 0.33 Closure (Stud) 0.76 0

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l TABLE 1 MAXIMUM FATIGUE USAGE FACTORS

• Component Material Fatigue Usage

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i Recirculation Outlet Nozzle Stainless Steel 0.10 Low Alloy Steel 0.30 Recirculation Inlet Nozzle Sh .376 Type 316 0.23

)

SA-312 Type 304 0.15 Low Alloy Steel 0.03 Feedwater Nozzle SA-105 GR II 0.89**

Core Spray Nozzle Stainless Steel 0.01

)

Low Alloy Steel 0.02 CRD Hyd. Ret. Nozzle Low Alloy Steel 0.36 Control Rod Drive Nozzle Ni-Cr-Fe Alloy 0.01 2" Instrumentation Nozzle Ni-Cr-Fe Alloy 0.06

)

l Support Skirt Low Alloy Steel 0.55 Refueling Containment Skirt Low Alloy Steel 0.33 -

Shroud Support Low Alloy Steel 0.17 Closure Low Alloy Steel 0.00

)

, Stud 0.76 Vessel Shells Water Level Area 0.01 1

Lower Head Area 0.03 O

• Fatigue Usage taken from Reference 3

• • Fatigue Usage taken from Reference 4

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O 4.0 ALLOWABLE CYCLE CALCULATION C) ..

The second task was to determine the allowable number of additional SRV blowdown cycles for the most limiting component. This was accomplished by evaluating the fatigue usage margin for each of the five' components selected in Section 3.0. Using this margin the allowable O

number of additional cycles can be calculated.

For each component, except the feedwater nozzle, the fatigue usage margin was based on the original Code design stress report. The O

feedwater nozzle fatigue usage margin was based on the feedwater nozzle modification stress report (Reference 4). This margin was calculated by subtracting the fatigue usage reported in Table 1 from the allowable fatigue usage of 1.0. The margin represents the amount of additional O

incremental fatigue usage the component can accumulate without exceeding the fatigue usage limit.

, Using the stress ranges available in the modification stress report v

in conjunction with the f atigue usage margin, the allowable number of additional SRV blowdown cycles was then determined. The additional -

cycles are shown in Table 2 for each of the five components.

O TABLE 2 ALLOVABLE NUMBER OF ADDITIONAL SRV CYCLES Component Fatigue Usage Cycles

.O I -

Margin Feedwater Nozzle 0.11 114 CRD Hydraulic Return Nozzle 0.64 325 Support Skirt 0.45 135

,0 Refueling Containment Skirt 0.67 269

! Closure 0.24 428 l

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'C Since the Feedwater Nozzle was limiting the results of this fatigue

{ assessment are shown in Appendix A.

1 4-1 O

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5.0 CONCLUSION

S q

The results in Table 2 show that the feedwater nozzle is governing since it is affected the most by the additional cycles.

Therefore, a total of 114 additional SRV blowdown events can be q'

tolerated in the 40 year life of the plant. It should be noted, however, that this analysis includes the following assumptions:

1. The assumed transients resulted'from a Safety Relief Valve (SRV) blowdown leading to complete depressurization. In reality, not all the SRV transients lead to this severe of a a

pressure drop. Therefore, the analysis is conservative since an additional number of cycles could be justified if the k O actual operating history were evaluated. Furthermore the results are not significantly affected by reasonable changes in the actual cooldown rate during blowdown. For example, even if the cooldown rate is 125*F/hr instead of the assumed O 100*F/hr, the event could still be treated as a typical SRV blowdown event and the results of this analysis would still apply.

.O

2. The additional cycles were calculated using the maximum alternating stress intensity for either the startup, turbine roll, or SRV blowdown event. Therefore, conservatism is built in thfs analysis.

Based on the limiting results for the Feedwater Nozzle a total of 114 additional SRV blowdown cycles can be allowed during the 40 year

O j plant life without exceeding Code fatigue usage limits.

'O 5-1

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6.0 REFERENCES

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1. " Reactor Thermal Cycles - Reactor Vessel," General Electric, San Jose, California, May 1967, (Drawing No. 729E762).

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2. " Nozzle Thermal Cycles - Reactor Vessel." General Electric, San Jose, California, May 1967, (Drawing No. 135B9990).

) 3. " Stress Analysis for the Peach Bottom Units 2 and 3 Reactor Pressure Vessel " Babcock & Wilcox Company, Mt. Vernon, Indiana, September 1970, GE Order No. 205-B1156, B & W Contract No.

610-0139-51, (VPF No. 1896-142-1).

)

i

4. " Stress Analysis for the Peach Bottom Units 2 and 3 Reactor i Vessel-Feedwater Nozzle," Ceneral Electric, San Jose, California, 1

August 1979, (22A6647).

c) 3 l

C 0

0 6-1 O

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D APPENDIX A ADDITIONAL CYCLES FOR THE FEEDWATER NOZZLE q The additional cycles calculation wrc based on an inadvertent

" Safety Relief Valve (SRV) Blowdown event, which bounds the actual plant transients. As a result of an SRV blowdown, a startup and turbine roll will also occur. Therefore, these two events were included in the additional cycles calculation.

q

" This calculat{onfatigue stress report was based on the feeawater usage analysis.

nozzledescribed The cases modification in Table A-1 were analyzed in this strass report. Three cases are related to the SRV blowdown event, two turbine roll cases (2 and

3) and the SRV blowdown case (8). The fatigue usage was calculated at the locations shown in Figure A-1; a summary of the n results are shown in Table A-2. The limiting fatigue usage

'J calculation is at Location 126 and is show in Table A-3.

The two limiting cases were the turbine roll cooldown at .14 minutes (TRCD.14) and the turbine roll warmup at .14 minutes (TR.14). It is seen that the SRV blowdown event alone has a stress amplitude of 7 ksi and does not contribute to the fatigue

'q usage. Even if this stress is somewhat higher due to a higher cooldown rate the stress amplitude would still be within the endurance limit of 25 ksi and the allowable number of cycles would be unchanged.

The objective was to determine the number of cycles that can be O added to the TRCD.14 and TR.14 without exceeding a fatigue usage of 1.0. The additional cycles were calculated using the following three step method:

1) The basic fatigue usage equation No. DesigD_ Cycles i 1.0 N . All w. Cycles 0 can be modified to determine the incremental usage. The Feedwater Nozzle incremental usage is calculated belows e Fat. Usage Limit - Total Usage = Incr. Allow. Usage 1.0 -

0.886 = 0.114 0 . 2) The allowable cycles for these two events are shown in Table A-3.

e Number of Allowable Cycles - Limiting Cases TRCD.14 1535 cycles

) TR.14 2911 cycles O

" Stress Analysis for the Peach Bottom Units 2 and 3 Reactor Vessel - Feedwater Nozzle," General Electric, San Jose, California, August 1979, (22a6647).

3 A-1

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3) The additional cycles are calculated as f ollows:

e Additional Cycles for the SRV Blowdown Event O .

Cycles + Cycles = Incr. Allow. Usage TRCD.14 Cycles TR.14 Cycles Cycles = (TRCD.14) (TR.14) (Incre. _ A11ow. _Usace)

() (TRCD.14 + TR.14) 114 cycles = (1535 cycl es) (2911_ cycles) (0.114)

(1535 cycles + 2911 cycles)

O Therefore, the limiting number of additional cycles for the SRV blowdown event are at least 114 cycles.

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A-2 O

t. _ . _ _ _ - - _ _ _ _ . _- . _ _ _ _ _ - _ . _ _ _ _ - _. .

O TABLE A-1 CASES ANALYZED IN MODIFICATIDN STRESS REPORT Case No. __ Code Description O .

1 ZEROLDAD Stresses equal zero 2 TRCD.14 Turbine roll cooldown 4 .14 min.

3 TR.14 Turbine roll warmup G .14 min.

4 TT126 Turbine trip at 25% power 4 126 min.

5 FHB64.5 Feedwater heater bypass 9 64.5 min.

O 6 LFP.14 Loss of feedwater pumps e .14 min.

7 LFP9 Loss of feedwater pumps 9 9 min.

B SVB1 Safety valve blowdown G 1 min.

9 NOOPM Normal operation - minimum conditions 10 HSB.14 Hot standby 9 .14 min.

11 SD102.73 Shutdown @ 102.732 min.

O 12 DSNHYDRO Design hydrotest.

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O TABLE A-2 FATIGUE USAGE

SUMMARY

Fatigue Fatigue 3 Location _ Usage _ Location _Usace_

85 0.15 198 0.0 90 0.01 211 0.39 121 0.74 216 0.0 126 0.89 235 0.31 3 151 0.20 240 0.06 156 0.01 259 0.20 175 0.14 264 0.08 180 0.01 289 0.34 193 0.05 294 0.0 0

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O TABL7. A-3 FATIGUE USAGE AS LOCATION 126 TOTAL USAGE = 0.89 O

TRANSIENT SALT NO. ALLOW NO. DESIGN FATIGUE TIMES (KSI) _ CYCLES CYCLES USAGE LFP.14 116. 435 130 0.30 DSNHYDRO O

ZEROLDAD 96. 716 62 0.09 LFP.14 ZEROLDAD 72. 1535 200 0.13 TRCD.14 O

ZEROLOAD 70. 1660 68 0.04 HSB.14 TR.14 56. 2911 436 0.15 HSB.14 O

LFP9 42. 7196 144 0.02 HSB.14 FHB64.5 36, 11113 270 0.03 HSB.14

-Q TT126 35. 13113 10 0.0 HSB.14 ,

NOOPM 35. 13620 1672 0.12 HSB.14 O

I NOOPM 31 18557 198 0.01 SD102.73 SVB1 7 >10. 56462 0.0 NOOPM

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