ML20073A220

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Executive Summary from Technical Rept 93177-TR-03,Rev 0 Re Plant Reactor Vessel Cumulative Usage Factor
ML20073A220
Person / Time
Site: Pilgrim
Issue date: 09/09/1994
From:
ALTRAN CORP.
To:
Shared Package
ML20073A210 List:
References
93177-TR-03, 93177-TR-03-R00, 93177-TR-3, 93177-TR-3-R, NUDOCS 9409200171
Download: ML20073A220 (11)
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Text

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EXECUTIVE

SUMMARY

Background and Purpose The reactor pressure vessel (RPV) at the Pilgrim Nuclear Station was designed in the late 1960's. Calculations for vessel life were based on fatigue analyses that were then performed by hand, using estimated numbers of cycles that were developed without the benefit of operating experience. The fatigue life of the vessel can now be calculated much more accurately by using modern analysis techniques and actual operating experience.

In recent years, it has become apparent that some of the original estimates for operating cycles that were used in the vessel design were low and in some instances would be exceeded prior to the 40 year design life. This project was undertaken to account for these increased cycle projections. Its primary purpose is to supplement the original analysis, remove unnecessary conservatism, and produce a more accurate calculation of vessel fatigue life using current analytical methods and current estimates for cyclic loading due to thermal, pressure and rnechanical loads.

A second objective of this project was to review and update the existing analysis of the feedwater inlet sparger. The original sparger analysis illustrated that the sparger may have to be inspected six times during the remaining plant life because possible leaking and deterioration of the sparger seals may lead to high fatigue stresses in the feedwater nozzle. This project included a review and revision of the sparger analysis to extend its useful life.

In addition to these primary tasks, the project included a related study to determine the effects of reducing feedwater inlet temperature on the fatigue life of the feedwater nozzles, ASME Section XI computer models, and preliminary screening criteria. The results of these studies are not included in this summary.

Critical Points for Fatigue Evaluations Since fatigue stresses are highest at points of structural discontinuity, the important locations for evaluation are easily identified. In the original analysis, the following locations were evaluated:

l 1. Closure region

2. Bottom head and support skirt
3. Feedwater nozzle (revised in 1982 to include rapid cycling) l 4. Steam outlet nozzles l 5. Recirculation inlet nozzles l
6. Recirculation outlet nozzles
7. Core spray nozzle
8. Vessel shell
9. 4" vent nozzle
10. 2" instrument nozzles f I a n .w.

94o9200171 94o909 PDR ADOCK 0500o293 P PDR

I1. 6" instrument spray nozzles

12. Drain nozzle
13. CRD nozzle
14. Jet pump instrument nozzles
15. Miscellaneous internals and attachments A detailed review of the original analysis confirmed that the assumptions and resUlts for items 11 through 15, although conservative, were acceptable; therefore these were not reevaluated in this project. Reanalysis of the vessel was performed for locations 1 - 11 above, which are shown in Figure 1.

Reanalysis for the feedwater nozzle (Item 3) was done primarily for rapid thermal cycling. The system transient analysis had been redone in 1979 and was acceptable; however, as part of this program, the feedwater sparger and nozzle was re-analyzed to include the more accurate cycle-count data that is now available.

Improvements to Original Analysis The original analysis was supplemented and revised to include the following three improvements.

1. Method of Analysis The original structural analysis was done by hand using interaction and seal-shell analysis. This type of analysis usually produces accurate results for stress near a discontinuity, but cannot predict the localizM peak stresses that are necessary to calculate fatigue life. As a result, stress concentration factors were multiplied by the calculated stresses to produce the peak stresses for the fatigue analysis. Good engineering practice required that the stress concentration factors be conservative, and these conservatisms were reflected directly into increased fatigue stress and consequent decrease in predicted fatigue life.

In addition, the original analysis could not predict stress gradients near structural  ;

discontinuities. 'Ihis limitation required that any flaw analysis that might be required in l these areas be based on maximum stress, rather than actual stress at the location of the flaw. (Flaw evaluation is not covered in this report, but related tasks involving flaw analysis were done as part of this overall program and are reported in Altran Technical l

Report No. 93177-TR-04).

I l

The method used for the ruvised analysis of the RPV uses finite element computer models to calculate peak stress directly and thereby eliminates the previous conservatism related to the use of stress concentiation factors (although Code allowable stress concentration factors were applied selectively to compensate for specific analysis uncertainties). It also provides accurate stress values in areas of rapid stress change, so that evaluation of ASME Section XI daws can be done more accurately, if required.

  1. 3t 77 GesesA
2. Cycle Counts Based on Operating Experience The cycle counts used in the original analysis were estimates made with very limited operating experience. The revised analysis uses cycle counts based on 21 years of PNPS experience, extrapolated to the 40-year design life of the plant. The extrapolation is based on a linear regression analysis plus three standard deviations to account for the data spread in the operating experience. A comparison of the original and new cycle counts follows.

Cycle Counts Original Revised Event Analysis Analysis Boltup 123 22 Hydro 130 22 Cold Startup 120 212 Hot Standby Startup 120 337 50% PWR Reduction 14,600 379 Loss of FW Heaters 80 10 Loss of FW Pumps 10 26 Turbine Generator Trips 40 27 Other Scrams 147 132 Full PWR Recire S/U 5 16 PWR Reduction to Hot Standby 118 176 Shutdowns 118 145 Safety Valve Blowdown S/D 2 47 Refueling Floodup S/D 118 20 Unbolt 123 22

3. Stress Range for Each Event in most cases, the original analysis made the very conservative assumption that every design cycle produced the maximum range of stress, regardless of the event. In fact, most events produce a stress range well below the maximum.

The revised analysis accurately accounts for the actual stress range for each event, then uses that stress with the cycle count for that same event to determine its incremental contribution to total fatigue.  ;

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Results The revised fatigue analysis showed substantial increases in fatigue life when compared to the original analysis. All sections of the vessel are acceptable for the full 40-year design life of the i RPV, including the feedwater nozzle and sparger. Fatigue Usage Factors that include the I increased cycle counts are all well below the limiting value of 1.0 as follows
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Fatigue Usage Factors Original Revised Comnonent Analysis Analysis Closure Region .77 .049 Closure Studs .79 .07 Bottom Head and Supt. Skirt .309 .044 Feedwater Nozzle System Transients .545m .637 Combined (Rapid and System) .678mm < . 8' Steam Outlet Nozzle m m Recirculation Inlet Nozzle .97 .037 Recirculation Outlet Nozzle .751 m Core Spray Nozzle .437 .01 Vessel Shell .435 .012 Vent Nozzle m m m m Instrument Nozzle

  • From 1979 analysis, using finite elements
  • Meets exclusion rules for ASME Code - fatigue analysis not required
  • This usage factor requires that the sparger be refurbished six times during remaining plant life.
  • Based on no refurbishment. i The large reductions in most of the fatigue factors are primarily the result of two factors: a

, substantial reduction in peak stress, and an accurate representation of the actual stress range of each event. The revised cycle counts show that both of the most severe transients (cold startups and loss of feedwater pumps) increased from the number projected in original analysis, but these increases were offset by the stress and stress range reductions resulting from this analysis.

l The increase in feedwater nozzle life is a result of eliminating several conservatisms contained j in the original analysis. Improvements in the revised analysis (relative to the original analysis) l included 'use of more accurate values for loss coefficients at the feedwater sparger seals and l gaps, coefficient of flow friction, coefficient of thermal expansion, and material corrosion rates. I In addition, many different evaluations were performed to account for possible variations in the positions of the feedwater sparger.

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The low fatigue usage factors resulting from the revised analysis show clearly the reactor vessel at PNPS can be operated safely for its full 40-year design life, and still maintain a substantial capacity beyond the 40 year design life, without refurbishment of the feedwater inlet sparger.

The maximum usage factor for the vessel is calculated at the feedwater nozzle and is <0.8.

The revised analysis also provided valuable information regarding stress gradients in the maximum stress regions. This information will be useful if ASME Section XI flaw evaluations are required in the future. Selected plots showing these stress and thermal gradients are presented in Figures 2 through 6.

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Figure 1 Re-analyzed Pressure Boundary Components of the Pilgrhn Reactor Vessel  ;

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ANSYS 4.4A JUN 24 1994 16:58:30 PLOT No. 2 POST 1 STRESS STEP =1 ITER =10 SY (AVG)

S GLOBAL DMX =0.760003 SMN =-32437 SMNB=-45128 SMX =48354 SMXB=57018 ZV =1

  • DIST=51.778
  • XF =118.724
  • YF =6.122 g -32437 g -23460 g -14484 g -5507 g 3470 g 12447

, , 21424 g 30400 m 39377 48354 f

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Heatup+ Bolt-up+ Pressure Figure 2 Closure Region - Displaced Geometry i and Sy Stresses for End of Heatup COLOR PLOT A W4M 7 .n n -

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ANSYS 5.0 A APR 14 1994 -

l 10:12:15 PLOT NO. 1 l NODAL SOLUTION l STEP =1 l SUB =1 l TIME =1 l SINT (AVG)

DMX =0.066031 SMN =1921 SMNB=1390 SMK =23581 SMXB=24896 g 1921 g 4328

" 11548 13954 1 16361 18767 g 21174 23581

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PILGRIM RECIRC. OUTLET NOZELE, PRESSURE =1000 PSI Figure 3 COLOR PLOT Recirculation Outlet Nozzle - Stress Intensity for 1000 psi Pressure Lead t

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3 ANSYS 4.4A JUL 22 1994 14:23:01 PLOT NO. 1 i POSTI STRESS

! STEP =1 ITER =1  :

TEMP SMN =71.417 SMX =545.952 ZV =1

  • DIST=17.205
  • XF =7.195 l *YF =184.539 EDGE g 71.417 l

" 124.143

" 176.869 229.595 282.322 g 335.048

, 387.774 g 440.5 g 493.226 545.952 1

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i IMMERGENCY SEUTDOWN Figure 4 COLOR PLOT Core Spray Nozzle - Emergency Shutdown Thermal Distribution at 0.05 Hour l

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MAY 20 1994 l 09:33:01 PLOT NO. 1 NODAL SOLUTION STEP =2 SUB =1 TIME =0.18 BFETEMP (AVG)

DMX =0.621327 SMN =373.053 SMI =528.334 g 373.053 g 390.306 j 407.559 424.813 l g 442.066 )

E 459.32 i

' ' 476.573 '

g 493.827 g 511.08 l 528.334  !

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e PILGRIM RECIRC. INLET NOZZLE, BLOWDOWN TRANS.

l Figure 5 COLOR PLOT Recirculation Inlet Nozzle -

Blowdown Transient Temperature Distribution at 0.18 Hour 10 -,n -

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ANSYS 4.4A l DEC 22 1993 8:08:56 ,

PLOT No. 5 '

POST 1 STRESS j STEP =4 ITER =1 SI (AVG) i DMX =C.766155 '

y SMN =2255 l

SMNB=468.117

( , SMX =49534 SMXB=59439 EV =1 DIST=90.75 XF =59.795 YF =-38.5 g 2255 g 7508 g 12761 g 18015 g 23268 g 28521

, , 33774 g 39027 g 44281

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HEAT-UP, PRESSURE + THERMAL Figure 6 COLOR PLOT Bottom Head and Support Skirt -

Stress Intensity at End of Heat-up git 11 =m -

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