ML20056E233
| ML20056E233 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/31/1993 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20056E232 | List: |
| References | |
| A-MECH-ER-009, A-MECH-ER-009-R00, A-MECH-ER-9, A-MECH-ER-9-R, NUDOCS 9308230010 | |
| Download: ML20056E233 (100) | |
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EVALUATION OF TIIEItMAL STItATIFICATION EFFECTS t
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i- ON TIIE ,
SIIUTDOWN COOLING LINE FOlt AltKANSAS NUCLEAll ONE, UNIT 2 AUGUST,1993 l'repared by
, AllIl Comljustion Engineering Nuclear Services Windsor, CT Allll Combustion Engineering Iteport No. A-MECII-Elt-009, Iter. 00 i
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9308230010 930817
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PDR ADOCK 05000368 f- gygy G PDR ts ASE A BnOWN BOVER:
TABLE OF CONTENTS i
i Section Title Page I
1.0 INTRODUCTION
. . .2 2.0 PROGRAM . . . 3
3.0 CONCLUSION
S . . .6 l
4.0 REFERENCE . . . . 7 LIST OF FIGURES Figure 1: SDC Line Measurement Locations - ANO2 . 8 Figure 2: ANO2 Shutdown Cooling Transient . 9 APPENDICES A. A-MECH-93-012." Return to Power of Arkansas Nuclear One-Unit 2 from Outage 2P-93-1." from B.T.Lubin ( ABB-CE) to R. Lane ( ANO2), May 11.1993 H. ABB-CE Calculation No. MISC-ME-C-164 Rev. 01, " Thermal Analysis of the ANO2 Shutdown Cooling Line, including Effects of Thermal Flow Stratification."
C.
ABB-CE Calculation No. A-MECH-CALC-023, Rev. 00, " Preliminary Stress and Fatigue Analysis of Thermal Stratification in the ANO2 Sloutdown Cooling Line
! 2CCA-25-14."
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1 ABB-CE Report No. A-MECll-ER-009. Rev.00 Page 1 of 9
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1.0 INTRODUCTION
ANO2 informed ABB-CE Nuclear Services that top-to-bottom wall temperature differences of 345'E were being recorded at Imcation I during the current mid-cycle cooldown of ANO2 (see Figures 1 and 2).
Based on a request by ANO2, an evaluation of the effect of this strati 0 cation on the stresses and fatigue life of the SDC line was performed. The objective of this evaluation was to support a return to power of ANO2 from this mid-cycle outage (2R9) and operation at least until the next scheduled outage in mid-1994 (2R10).
This request focused in three speci6c areas:
- 1. Identification of regions for subsequent NDE examinations and justi6 cation for the need for such examinations before the next scheduled outage.
- 2. Identification and justification for additional thermocouple placements to obtain further data on the magnitude and distribution of the top-to-bottom wall temperature differences.
- 3. Evaluation of the stress levels and fatigue usage due to the recorded temperature differences.
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The proposed program (Reference) included two Tasks:
. Task 1 Preliminary Evaluation: a letter report supporting return to power, with reliance on current NDE and thermocouple data This letter report was based on an engineering-assured. preliminary stress and fatigue evaluation. (A copy of the letter report is included as Appendix A. )
. Task 2 Final Evaluation: this report, which was generated to formally document the Quality Assured analyses that were performed and the conclusions that were reached.
ABB-CE Report No. A-MECll-ER-009, Rev.00 Page 2 of 9 J
2.0 PROGRAM !
The objective of the program was to support the return to power. This can be achieved by demonstrating that stress levels are acceptable and that the fatigue life of the line has not been significantly shortened by the presence of stratification. '
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Examination for evidence of cracks (NDE), and placement of additional thermocouples are both related to the evaluation of stress and fatigue. NDE is necessary at locations of high stress and/or high fatigue and the distribution and magnitude of wall temperature influences ,
. the location and magmtude of the stresses. !
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The fatigue usage factor is a function of the magnitude of the stresses and the number of !
stress cycles the line has experienced, both since the entering initial service and since the last i NDE of the SDC line in outage 2R8. The value of usage factor will indicate the design !
margin remaining in the line, and the rate at which fatigue damage is being accumulated. l With a small increase in usage factor since the last NDE, it is reasonable to assume that l significant damage to the line will not occur before the next scheduled outage. i f
The evaluation is based on the maximum temperature difference of 350*F and assumes that i the entire horizontal portion of the line from Elbow B to Elbow F (see Figure 1) experiences i the maximum temperature difference.
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2.1 THERMAL STRESS ANALYSIS l The thermal stress analysis is documented in the recorded calculation in Appendix B. l
. Forces, moments, support loads, displacements and stresses for the stratified conditions were !
calculated using the SUPERPIPE computer code, based on the following set of assumptions: f
- 1. Tributary lines of .50",1.0" and 3" were ignored.
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- 2. The model was terminated at the first elbow beyond support H15. i
- 3. Spring hangers were assumed not to have reached their travel limits.
f 4 Line temperatures were assumed, as follows: ;
I Case a: Stratified Flow !
t A-B 560*F. Uniform l B-F 560*F Top, 210 F Bottom i F-G 80*F. Uniform _
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Case b: Uniform Expansion ;
A-G 500*F Uniform ;
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The uniform expansion case was included as a possible enveloping case for other .
, thermal loading conditions which may have been part of the original design basis. l:
- 5. The influence of thermal stratification can be included using the results for line !
rotation extrapolated from the ANO2 surge line analysis.
The temperature distribution assumed in the analysis is based on measurements taken at i ANO2 (e.g. Figure 2). For the shutdown cooling line, being closed at the valve, the j variation in top to bottom difference in wall temperature is due to changes in secondary flows !
(" turbulent penetration") in the lines. Preliminary indications are that the fluid is thermally {
stratified without the presence of any interface; region of rapid changes in the temperature j distribution of the fluid. This differs from the surge line in which changes in operating !
conditions results in the propagation of fluid relative to a stationary layer. Based on the l absence of this hot to cold fluid interface variations in fluid and thus wall temperatures at the !
interface location (" striping") has been neglected. l l
The thermal stress analysis determined that the highest stressed location for the stratified l conditions was the first elbow directly below the Hot Leg (Figure 1: Elbow B); this l determination was based upon the relative ranking provided by the SUPERPIPE computer ;
code. i 2.2 STRESS AND FATIGUE EVALUATION l l
1 The stress and fatigue evaluation is documented in the recorded calculation in Appendix C. !
The evaluation calculated stress ranges which resulted from the thermally stratified conditions !
in conjunction with the two design basis transients which contributed most significantly to the !
usage factor of the elbow. These stress ranges were then compared to ASME Code i allowable and it was determined that the ranges were within the limits provided by the Code.
Additionally, a fatigue usage factor of 0.073 was calculated for Elbow B. This number is the sum of two values: 1) 0.044, the design basis usage factor for the location and 2) 0.029, which accounts for 350 .1T thermal stratification in the SDC Line during 68 Heatup- l Cooldown cycles and an estimated 436 Power Reduction, Reactor trip, and Loss of Coolant
. Flow transients.
A fatigue usage factor of 0.113 was calculated for the stress-limiting elbow for an additional l 10 Heatup-Cooldown cycles and 1000 additional Power Reduction, Reactor trip, and Loss of j Coolant Flow transients.
it was noted in the fatigue analysis that the stress-limiting location did not have the highest usage factor for the entire SDC Line. The location on the Line with the highest usage factor !
(0.398) was evaluated for the effects of the additional cycles. It was determined that the j usage factor for this location would become 0.467 (by applying the increase in fatigue for the- l stress-limiting cibow to this location). Thus, for either the stress-limiting location or the i fatigue-limiting location, the resulting usage factor is well below the ASME Code limit of 1.0. jl ABB-CE Report No. A-MECH-ER-009, Rev.00 Page 4 of 9 l
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i This analysis has been completed based on the largest variation in wall temperature observed !
l to date being less than 350 F. The maximum temperature difference possible in the first !
- ' horizontal run is directly proportional to the Hot Leg temperature (-565 F) less the ambient i temperature in the first horizontal run (-150 F is the lower temperature observed at the i bottom thermocouple). While it is possible that a larger AT could be experienced, it is not i
expected to have significant effect on the total usage factor and thus the number of additional cycles. This is based on the low usage factor observed to date (and near future) and due to !
significant conservatism used in this analysis (i.e., using rigid support stiffness, maximum !
l aT used for even minor power reductions, SDC first horizontal run considered stratified the i
. entire length, etc.).
I Both the thermal stress analysis and the stress and fatigue evaluation were performed j j* according to Section III of the 1986 ASME Boiler and Pressure Vessel Code. The use of the ;
- 1986 Code is based on the recommendation in NRC Bulletin 88-11 on Pressurizer Surge Line Thermal Stratification to use the latest ASME Section III requirements for high cycle fatigue.
The use of the 1986 Code is consistent with the analyses submitted to the NRC as part of the ,
) response to NRCB 88-11 on " Pressurizer Surge Line Flow Stratification Evaluation " (CEN- ;
1 NPSD 546-P.Rev l-P) by the Combustion Engineering Owners Group. ;
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3.0 CONCLUSION
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Design Analyses were completed that evaluated the effect of top-to-bottom wall temperature i differences of 345*F on the SDC line recorded May 4,1993 during the reduction in power l for the 2P-93-1 mid-cycle outage. These analyses support the conclusions that: ,
- 1. Based on the 1986 ASME Code, the accumulated fatigue usage factor is below the ;
Code limit of 1.0. Furthermore, the anticipated number of additional thermal cycles !
will not increase the total usage factor above the ASME Code limit of 1.0. i
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- 2. The highest stress levels due to thermal stratiGcation are calculated in the elbows
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closest to the hot leg which were subject to NDE during the 2R8 outage. The increase in fatigue life usage will not exceed the ASME limit for a limited number of ;
additional thermal cycles. Thus, the present NDE results can be used as evidence of the present status of these components and the SDC line should not require additional !
NDE prior to a Return to Power. i
- 3. The addition of more thermocouples to the SDC line prior to the 2R10 outage is not !
necessary. Additional thermocouples would provide information on the extent of j strati 6 cation within the line and would allow a rennement of the stress analysis, !
leading to a possible reduction in calculated stress levels. However, based on the !
present conservative assumptions, the fatigue usage factor is suf6ciently low that l additional margin would not be needed for a significant number of additional thermal cycles.
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I 4.0 REFERENCE !
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C-E Proposal No. 93-241-63A, " Evaluation of Thermal Stratification Effects in the i Shutdown Cooling Line on Return to Power of Arkansas Nuclear One-Unit 2," dated May, l 1993. ;
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APPENDIX A 1
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. May 11,1993 A MECII.93-012
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Mr. Rick Lane, Manager Mechanical, Civil and StructuralDesign
. Entergy Operations, Inc.
Route 3, Box 137C Russellville, AR 72801 SUBJECr: Retum to Power of Arkansas Nuclear One-Unit 2 from Outage 2P 93-1
Dear Mr. Lane:
Attached is a report provided by ABB-Combustion Engineering summarizingthe effort to-date in support of the Return to Power and operation from outage 2P 93-1, of Arkansas Nuclear One-Unit 2. The work represents the preliminaryevaluation of the stressand fatigue life in the Shutdown Cooling line following the recording of top-to-bottom diffennees of wall temperatures of up to 345'F during cooldmvn.
This work supports the following conclusions:
- 1. Eased on the 1986 ASME Code the accumulated fatigue usage factor is below the Code limit of 1.0. Furthermore, a number of additional thermal cycles will not incmase the total usage factor above the ASME Code limit of 1.0.
- 2. The highest stresslevels due to thennal stratificationare calculated in the elbows closest to the hot leg which were subject to NDE during the 2R8 outage. The increase in fatigue life usage will not exceed the ASME limit for a limited number of additional thermal cycles. Thus, the pnsent NDE results can be used as evidence of the present statusof these components and the SDC line should not require additional NDE prior to a Return to Power.
- 3. The addition of more thennocouples to the SDC line prior to the 2R10 outage is not necessary. Additional thermocouples would provide information on the extent of stratificationwithin the line and would allow a refinement of the stress analysis, leading to a possible reduction in calculated sinssievels. However, based on the present conservative assumptions,the fatigue usage factoris sufficiently low that additional margin would not be needed for a significant number of additional thermal cycles.
ABB Combustion Engineenng Nuclear Power
- a. . . ~.o,_ ,s u . m ., _ ne f, s.,r,,,,,, m o,m n P%f ( Ptef^f* hia 'n M I b a {2D3) 2'db 'j. '
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. This evaluation, while being engineering assund,should be considered preliminaryuntil completion of formal Quality Assurance in compliance with ABB Combustion Engineering Nuclear Services QAM-100 procedures. This final report will be completed as a prugram being proposed to ANO2.
. Should there be any questions on these results,please call me at 203-285-4996.
Sincerely, ABB COMBUSTI N-ENGINEERING WCLEAR SE IC I Ba T.Lu n
/A SupeMsor, Fatigue Evaluation Senices xc: M.S. Mcdonald (ABB-CE)
R.W. Bradshaw(ABB-CE)
W.A. Goodwin(ABB-CE)
D.F. Balsley(ABB-CE)
R.C. Sykes(ABB-CE,ANO2 RSSM)
W. Greeson(ANO2)
. J. Martin (ANO2)
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SUMMARY
AND CONCLUSIONS IN EVALUATION OF TIIERMAL STRATIFICATION IN TIIE SIIUTDOWN COOLING LINE ON TIIE RETURN TO
- POWER OF ARKANSAS NUCLEAR ONE-UNIT 2
, llACKGROUND: On May 4,1992, ANO2 Informed Nuclear Services that top-to-bottom ,
wall temperature differences of approximately345*F had been reconfed during the power reduction for the mid-cycle outage of ANO2. These values are above the 300*F differential temperature previously reconfed at Location 1 (shown in the attached Figure). Temperatures at Location 2 were close to ambient (80*F) and were uniform !
around the line. In addition,it was noted that these temperature differences were related to reductions in power, increasingin magnitude with larger percent reductions in power.
A request was made by ANO2 to pruvide assistanceln evaluating the effect of this stratificationon the stressesand fatigue life of the SDC line. The objective of this evaluation was to support a Return to Power of ANO2 from this mid-cycle outage (2P-93-1) and to support operation for a limited number of addidonal thermal cycles. ,
This request focused in three specific arras: *
- 1. Evaluation of the stresslevels and fatigue usage due to the reconled
, temperature differences, with allowances for additional future thermal cycles.
- 2. Identification of regions for NDE examinations.
- 3. Identification of locations for additional thermocouple placements. ;
The purpose of this report is to document this preliminaryevaluation and the conclusion that the presence of thermal stratificationin the shutdown cooling line has not, and will ;
not, for an additional number of thermal cycles, cause the SDC line to exceed ASME i code limits on fatigue, and that ANO2 can return to power and operate safely.
STRESS ANALYSIS: The purpose of the stressanalysiswas two-fold: to determine the
- locations of the maximum stress for identifying NDE locations and to calculate the fatigue usage factor for the line, from the initial plant criticalityto the end of the current ,
fuel cycle.
i This analysiswas done using the SUPERPIPE code, utilizing experience gained in analysisof pressurizersurge lines. The analysisconservatively assumed that: the SDC line is anchored at the RCS nozzle, top-to-bottom wall temperature differences are assumed constant over the upper horizontal portion of the line (Figure;B-F), with a temperature difference of 350*F, while the remainder of the line is taken at a uniform temperature of 560*F in the vertical section below the RCS nozzle (Hgure;A-B) and 80*F at the vertical run before (Figure; F-G) and horizontal run (Mgure;G-H) before
and after the isolation valve. The effect of the thermal stra . 41s accounted for
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using results for line rotation based on the surge line analyst. .se for a top-to-bottom temperature difference of 350*F.
A fatigue usage factor of 0.13 was computed for the limiting elbow (F1gure; Elbow B) !
, based on an estimated number of cycles during which stratification!n the SDC line is assumed to have occurred since initlat criticality.
1 In addition, to support continued safe operation, future thermal cycles must be accounted for. For the number of cycles specified below, an incrementalincmase in ]
usage factor of 0.12 would occur, resultingin a total usage factor of .25, which is below the ASME code allowable of 1.0.
Based on this analysis, continued operation can proceed for the number of SDC thermal stratificationeycles within the SDC line wall temperature differences listed below.
Measured wall temperature differences greater than 350*F will require a re-evaluation of ,
sinssesand of the usage factor.
Wall Temperature Difference Number of Cveles i
> 350*F Re-evaluation of usage factor ;
200"F - 350"F 1000 SDC thermal stratification !
cycles i
< 200*F Unlimited SDC thermal stratification cycles ,
These calculations,done for the elbow below the hot leg nozzle (Figure; Elbow b) which
, showed the highest stresslevels due to thermal stratification,were based on methods and procedures required in Section III of the 1986 ASME Boller and Pressure Vessel Code. The use of the 1986 Code is based on the recommendation in NRC Bulletin 88-11 on PressurizerSurge Line Thermal Stratificationto use the latest ASME Section III requimments for high cycle fatigue. The use of the 1986 Code is consistent with the analyses submitted to the NRC as part of the response to NRCB 88-11 on " Pressurizer Surge Line Flow Stratificatloa Evaluation " (CEN-NPSD 546-P,Rev I-P) by the Combustion Engineering Owners Group.
These results support the conclusion that the margin In fatigue life of the SDC line is sufficiently high to allow the plant to return to and continue safe operation.
NON-DESTRUCTIVE EXAMINATION: The resultsof the SUPERPIPE analysisshow :
that the maximum simsslocations am at the cibows closest to hot leg nozzle. ;
Based on information pmvided by ANO2, Non-Destructive Examinations at these locations and the weldolet region at the hot leg injection line junction, wen performed
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during outage 2RS. The results showed no evidence of flaws Indicative of cracks. In addition, the UT techniques utilized in this ISI were enhanced in accordance with recommendations in NRC Ilulletin 88-08, Supplement 2. The inspection included
. enhanced UT of welded regions, visualinspection of the weldolet and UT Inspections of base metal in the elbows.
Results of the fatigue analysisIndicate that the additionalloss of fatigue life from the 2R8 through 2P-93-1 outages and for the additional thermal cycles specified above will be minor.Thus the NDE results taken during the 2R8 outage should be a sufficient Indication as to the condition of the high stressregions until the next scheduled outage.
ADDITIONAL TIIERMOCOUPLE LOCATIONS: The current SUPERPIPE analysis conservatively assumes that the top-to-bottom difference in wall temperature is uniform over the entim upper horizontal section of the SDC line. A more local distributionin the effect of the stratificationwould resultin a reduction in stmssvalues but is not expected to change the location of the maximum stresses. _
The usage factor based on the stressescalculated with the current assumptions,and based on the 1986 ASME Code, have a sufficient margin below the ASME code limit such that only a limited benefit would be gained from using a more accurate, less i
. conservative, distribution on wall temperature diffemnce.
CONCLUSIONS: The following conclusions are based on this evaluation:
- 1. The accumulated fatigue usage factor of 0.13 is well below the ASME Code limit of 1.0. Furthermore, the additional conservative increase in usage factor of 0.12 1 for a number of additional thermal cycles will not inemase the total usage factor above the ASME Code limit.
- 2. The highest stresslevels due to thermal stratificationare calculated at elbows closest to the hot leg which were subject to NDE during the 2R8 outage. The increase in fatigue usage factor will not exceed the ASME Code limits for the recommended number of additional thermal cycles. Thus, the current NDE results can be used as evidence of the present statusof these components without additional NDE prior to a return to power.
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- 3. The addition of more thermocouples to the SDC line prior to the 2R10 outage is not necessary. Additional thermocouples woald provide information on the ;
location of stratificationwithin the line and a refinement of the stress analysis ,
leading to a reduction in calculated stresslevels. IIowever, based on the present 1 conservative assumptionsthe fatigue usage factor is below the ASME Code limit j therefom this additional margin will not be needed for a significant number of additional thermal cycles. l
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Author S d. M s r trJ d Date 7 2 9 - 9.3 Calculation contains safety related design information: Yes / No VERIFICATION STATUS: COMPLETE The design information contained in this document has been verified to be correct by means of Design Review.
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A Page Number RECORD OF REVISIONS l.Lo . Date Sections Involved Prepared By Apprevals 00 (ol2sl43 ALL- ogiGwnL C.t2. Te w 0T #F. 6A 45LCY ISSWL Gli. hEEFEc5tcI I #7 f3 ALL flys 6. C. Avsvi,J b. F. cAA ut. e y 4A. po en rey g,e,
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Table of Contents Page No.
Purpose 4 Discussion 4 Assumptions 7 Results q Ref erenc es 22 Computer Code References 23 Computer Run Log 24 Body of Calc 1 Preliminary and Fin s1 Model Ceometry 25 H. Preliminary Analysis Stratified Flow Input 27
~1 Preliminary Analysis Boundary Conditions 32 D, Preliminary Analysis Su= mary 33 l
- 7. SDC Line Superpipe Model Diagram 34 l 9. Changes for Final Analysis 35
. ET Stratified Flow Input for Final Analysis 36 l @ Final Analysis Simnry 39 g SDC Line Superpipe Model Control Point Coordi- 40 nates Appendix - Run #5 input File Listing Al ABB Combustion Engineering Nuclear Power Form # 0090257-B (Rev. 7/90)
A RR 7%99 ll ]( ~/M & ( ~ N 00 Calculation Number Rev. Page Number PURPOSE - To analyze the ANO2 shutdown cooling (SDC) line with the Superpipe computer code and determine loads and Equation 10 & 12 (Ref. 5) stress ranges resulting,frem the analysis of the following two cases: 1) an arbitrarily selected bounding case involving only linear thermal expansion, and 2) a thermal flow stratification case in which stratification occurs in a portion of the SDC line. These cases are outlined in the discussion section and the body of the calculation. .
- DISCUSSION - The basic SDC line geometry was based on Ref. 1 & 2 (see pp. 25-26). The 8' line off of the SDC line in the vici-nity of support H9 was added to the model at a later point in the analysis (see pp. 34-35). This geometry was based on Ref. 8.
See Ref. 10 for a description of the method used for analyzing thermal flow stratification. See p. 34 for a diagram of the computer model. A listing of the input file used in the final iteration run of the final analysis is contained in the appendix. Ref. 6 transmitted preliminary results generated by computer runs #1 & 2, in which flow stratification of delta 340 deg F was assumed along the first horizontal plane run off of the hot leg. *he rest of the line was only subjected to linear thermal expansion (see p.32). This analysis, the preliminary analysis, was run with the basic SDC line geometry described above. Subsequently, the model and the stratification case were ABB Combustion Engineenng Nuclear Power
- . n . w . .... m
JL D D i
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In/SC-InE-C- LW o( . Calculation Number Rev. Page Number modified per pp. 35-36. Computer runs #43 35 the final analysis, were made. The initial vertical section of the line off of the hot leg was at a temperature of 560 deo F, followed by the stratified section of the SDC line, which was subjected to a delta T of t 350 deg F, while the remainder of the line was held at a constant temperature of 80 deg F. Additionally, the vertical hanger (H1 & f H2) stiffnesses were changed to 1600 and 1200 #/in, respectively, and, as noted above, the 8' line off of the SDC line was added to the model. Therefore, runs #4-freflect the most accurate ) available model data, and the most complete SDC line modeling. It should be noted that the addition of the 8" line to the overall nodel did not significantly alter the results of the analysis, and that the change of vertical hanger stiffnesses i from 1500 #/in to their final values had a negligible effect
. en the analysis results. Additionally, the stratification case t
input that was used initially was the core severe of the two
.overall thermal inputs used, when considering both the stra-4 and unstratified portions of the line. In all of the computer i
runs made, the linear expansion only case was a constant 500 deg F, l i which is a conservative overall temperature to apply to the SDC ! i i 1 line. 4 9 Because some of the SDC line supports contain gaps in places (i.e., H9A & H9B), the final support configuration was arrived I at itere.tively in both the preliminary and final analyses. This ABB Combustion Engineering Nuclear Power , form # 00902$7 B <Rev. 7 901 . r
ARR 7% B B rn/v-rM-( - /6V Of Calculation Number Rev. b Page Number was done by assuming an initial set of open and closed support gaps and altering this configuration during successive computer runs until the final configuration led to a set of consistent support loads. If a vertical support gap should have been closed but was not during a particular iteration, the support was made rigid in the next iteration and the support point in the model was moved a distance equal to the gap size. More particularly, the support point was moved in the (+) direction if the support topped out and in the (-) direction if the support should have bottomed out during the iteration. This was done by altering the support point displacement in the stratified flow input section of the computer run input file. If, on the other hand, a gap previously taken to be closed opened, the support was given a negligible spring rate (1 W/in), and the support point displacement was restored to th6 original value in the next iteration. For the preliminary analysis, run 42 reflects the final support configuration, and for the final analysis, run #.( reflects the final support configuration) b e" e- Mh4h
- MME-Fci- f/te 500 ' ttw+#rno ca se. , rta s I eP 2 M yJed hr f) K fe V if1A h t< J y it % Ih 00' 4 s k le L uea er d e &~d - I
. ABB Combustion Engineering Nuclear Power Form e 0090257-8 IRev. 7;901 1
A It D 7%ERID tY)/S C-t M -( ~ 12 / oI
. t Calculation Number Rev.
7 i' Page Number It should be noted that although loads and stress range values are available from the analyses run, only loads should be used to assess piping stresses in an absolute sense. Getting accurate stress ranges (i.e. equn. 10 & 12 results) requires consideration of all loading conditions in the original design basis for the . subject line acting in combination with flow stratification. This analysis does not examine the necessary prerequisite conditions. However, a comparison of the magnitudes of the equn. 10 & 12 results along the SDC line can be used to determine locations at which the stresses are likely to be most severe. 1 ASSUMPTIONS - .
. Mvu>4 ^D
- 1) The (-) vertical gap sire at supports H9A & H9B is equal to 1/2*. Inspection of the Ref. 3 support drawing shows that this is a reasonable assumption, and the final results indi-cate that the downward motion at these supports is much <
the assumed limit.
- 2) The effects of 3* and smaller lines off of the 14* SDC i
line are negligible, and therefore these lines do not need to be included in the model. Comparison of runs #2 & 6) in which the model does not and does have the 8* line included indicates essentially the same behavior. Therefore, the inclusion of significantly smaller lines would have F ABB Combustion Engineering Nuclear Power Form a 0090257 B (Rev. 7.906 l
ABB __ Calcutstion Number Rev. 2 Page Number a negligible effect on the results.
- 3) The horizontal plane clearance at the H16 guide (Ref. 2) is sufficiently large to allow for no contact. This assumption f
is valid because the partial stratification case run herein is unlikely to cause a significantly larger movement at this location of the SDC line than some of the linear thermal expansion cases for which the support has been designed. Further, it is likely j l that the stratification analyzed is less severe than the actual l stratification experienced. [
- 4) The end of the model is fixed in all directions but axial translation. The true end condition of the portion of the overall SDC line modeled for this analysis is somewhere
~
between free in axial translation and completely fixed, with the more reasonable approximation being the one used in this analysis. Run #3 was made with the end completely fixed, and the resulting final support configuration along the rest of the model remained the same, indicating that using either end condition produces essentially the same results. j I
- 5) Vertical hangers H1 & H2 had-an assumed stiffness of 1500 #/in I in the preliminary analysis. Subsequent undocumented data indi-4
. cated that the stiffnesses were 1600 and 1200 #/in, respectively.
These values were used in the final analyses. All of the other 1 active supports were assumed to be rigid. The fact that the 4 l ABB Combustion Engineenng Nuclear Power i c _ ,nean ,= , ,... rn
1 l A45 guic-m&C- /M Calcutstion Number 00 Rev. l
. I k
Page Number changes made to the hanger stiffnesses in this calculation's runs had a negligible effect on the results indicates that the j
~
final values used will produce.results that are within the accuracy requirements of this analysis. Similarly, inspection of the support drawings validates the assumption that the other supports are rigid, since they only need to approach being rigid when compared to the stiffness of the SDC line in order for the analysis to produce sufficiently accurate ; results.
- 6) It was assumed that neither of the vertical spring hangers bottomed or topped out in the final support configurations.
In general, experience with surge lines hao shown that the maximum hanger displacements encountered in this analysis ;
.342*, +.376') are not sufficiently large to cause the !
( vertical hangers to top or bottom out. Therefore, the I assumption is validated. i FI.SULTS - The results are contained en pp. 11 - 21. They consist of j tabulated SDC line forces and moments in the Superpipe local coordinate system along with a SDC line diagram defining the local coordinate systems for the straight i 4 piping sections in the model. Units are inches and pounds. Therefore, the results on the straight pipe side of a l l straight piping / elbow interface are the results that i i ABB Combustion Engineenng Nuclear Power i r m._,--,.,...... ,~
ARR M ED D rnliC-ml=-C- E4 oI Calculation Number Rev. IO Page Number
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are consistent with the local coordinate system diagram. It should also be noted that to be totally consistent,
. the signs of the loads at the DCP ' coming out* of an elbow (e.g., DCP SB) should be reversed, while the signs of the loads
- going into* an elbow / piping juncture (e.g., SA) should remain as is. For example, the correct set of loads for 5A & SB for the final analysis stratified flow case (see p. 19) are as follows:
LOC MX MY MZ 5A - II7040fpo -18929433 - I44'l999. So SB -t2 gyp 3,g7 ;zo4ggy,g9 + f bg7/33,SO , l i l Results are presented for the constant 500 deg F and the l 1 ! I partial stratified flow cases for both the preliminary ' and final analyses. The results reported in this calculation are thermal only loads. The 500 deg F case (THMN) results are for information only i since there is no real basis for the computer code input for this case. Flow stratification results (SFL1) should be selected from run #2 & 6 outputs in the most conservative
- manner . It is reconunended that the larger of the FCJulti be, selected on a location by location and component of load by cornponent of load basis.
ABB Combustion Engineering Nuclear Power i sa. e oceo m a m n nos l
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l. ARKA!!SAS POWER & LIGHT SHUTDOWN COOLItM LItJE - PhELIMINARY A!JALYSIS PESULTS 500 DEG F CONSTANT TEMP CASE LOAD CASE tJ0. 3 ( THMll) , FORCES A!JD MOMEtiTS Ili LUCAL C00RDI!4ATES RUN SOP DCP AXIAL Y Z XX YY ZZ FORCE FORCE MOMEf3T MOMEliT MOMEt4T GROUP MMB t1AME FORCE (LB) (LB) (LB) (LB.IN) (LB.IN) (LB.1H) PRH1 *4 1 1 1705.00 0.00 644.17 226941.08 164990.05 -34202.02 2 17 0 5. 0.^ 0.00 644.17 226941.00 188225.67 -34202.02 1705 "', 0.00 644.17 226941.08 211471.39 -34202.02 3 4 1705.00 0.00 644.17 226941.08 234717.11 -34202.02 f)Cos t. /(ug) je2, SL 2A 1705.00 0.00 644.17 226941.08 257962.92 -34202.02 SR 2A 1705.00 -455.49 -455.49 226941.08 -158222.84 206591.81 6L 2B 455.49 1705.00 -455.49 168594.50 216569.42 178140.34 6R 2B 455.49 -1705.00 455.49 168594.50 -216569.41 -178140.34 7 3 455.49 -1705.00 455.49 168594.50 -212927.02 -164506.03 8L 4A 455.49 -1705.00 455.49 168594.50 -209235.56 -150688.25 BR 4A 455.49 -455.49 -1705.00 168594.48 150688.27 -209235.56 9L 4D 644.17 0.00 -1705.00 24032.65 198314.92 -204939.50 9R 4B 644.17 -1705.00 0.00 24032.64 -204939.50 -198314.92 10 644.17 -1705.00 0.00 24032.65 -204939.52 -157547.94 11L SA 644.17 -1705.00 0.00 24032.64 -204939.50 -116780.79 11R SA 644.17 0.00 1705.00 24032.65 -116780.79 204939.50 12L SB 455.49 455.49 1705.00 88199.11 -38130.77 200643.42 12R 50 455.49 -1705.00 455.49 88199.11 -200643.42 ~38130.77 13L H2 455.49 -1705.00 455.49 88199.11 -184188.47 234E3.39 13R H2 455.49 -1760.00 455.49 88199.11 -194188.47 23463.39 14L 6A 455.49 -1760.00 455.49 88199.11 -182212.91 31096.82 ' 14R 6A 455.49 -1760.00 455.49 88199.11 -182212.89 31096.82 15L 6B 1760.00 455.49 455.49 175298.48 95113.52 50899.38 ISR EB 1760.00 0.00 644.17 175298.47 31264.13 103246.73 16 H) 1760.00 0.00 644.17 175298.47 52218.02 103246.73 17 H4 1760.00 0.00 644.17 175298.47 57005.73 103246.73 18L 7A 1760.00 0.00 644.17 175298.47 65488.02 103246.73 18R 7A 1760.00 -455.49 -455.49 175298.50 -119314.04 -26698.87 19L 78 455 49 1760.00 -455.49 129685.70 164926.83 -56402.71 19R 7B 455.49 -1760.00 455.49 129685.69 -164926.81 56402.71 05376.48 ** 20L H1 455.49 -1760.00 455.49 129685.69 -157428.34 99 20R HI 455.49 -1499.00 455.49 129685.69 -157428.34 05376.48 21 VA1A 455.49 -1499.00 455.49 129685.69 -144093.38 129261.14 (7 22 455.49 -1499.00 455.49 129685.69 -133721.75 163393.56 'ik i
'1s 23 VAID 455.49 -1499.00 455.49 129685.69 -123350.09 197526.14 *-
24L 8A 455.49 -1499.00 455.49 129685.69 -123324.63 197609.88 iD 129685.70 197609.09 123324.63 bd 24R BA 455.49 455.49 1499.00 8 25L BB 0.00 644.17 1499.00 -58026.82 255568.23 112952.99 25R 8B 0.00 -1499.00 644.17 -58026.84 -112952.99 255568.25 26 H5 0.00 -1499.00 644.17 -58026.84 -111345.48 259309.02 0.00 -1499.00 644.17 -58026.84 -107154.70 269061.16 sl 27 H9A cs N F5 o O
. . . . . -- . . . ~ . . . .. . . - . . ,- . .
=
l-; ARKANS AS POWER & LIGHT SHUTDOWN COOLINd LINE - PREL1H114ARY ANALYSIS RESULTS 500 DEG F CONSTANT TEMP CASE LOAD CASE NO. 3 (THMN), FORCES AND HONENTS IN LOCAL COORDINATES (CONTD.) AX1AL Y Z XX YY ZZ RUN SOP DCP MOMENT GROUP MMB NAME FORCE FORCE FORCE HOMENT HOMENT (LB) (LB) (LB) (LB.IN) (LB.IN) {LB.INI PRN1 (CONTD.) 644.17 -58026.84 -96677.75 293441.53 28L *19 0.00 -1499.00 2BR H9 0.00 17642.99 644.17 -58026.84 -96677.75 293441.53 29L H98 0.00 17642.99 644.17 -58026.84 -86200.01 6489.44 29R H9B 0.00 -93.28 644.17 -58026.84 -86200.01 6489.44 0.00 -93.28 644.17 -58026.84 -63151.54 9827.01 30 VA2A -48483.84 11950.91 31 0.00 -93.28 644.17 -58026.84 32 VA2a 0.00 -93.28 644.17 -58026.84 -33816.10 14074.83 33L H17 0.00 -93.28 644.17 -58026.84 -16354.52 16603.29 33R H17 0.00 196.19 622.14 -58026.84 -16354.52 16603.29 34L 9A 0.00 196.19 622.14 -58026.04 -2188.24 12135.92 34R 9A 0.00 -622.14 196.19 -58026.83 -12135.93 -2188.24 35L 9B 622.14 0.00 196.19 7668.56 -53559.47 11978.04 35R 99 622.14 196.19 0.00 7668.56 11978.04 53559.47 36 H21 622.14 196.19 0.00 7668.56 11978.04 52921.28 622.14 196.19 0,00 7668.56 11978.04 45518.26 37 11978.04 18115.19 38 622.14 196.19 0.00 7668.56 39 622.14 196.19 0.00 7668.56 11978.04 30712.13 s 40 622.14 196.19 0.00 7668.56 11978.04 23309.07 - 41 H16 622.14 196.19 0.00 7668.56 11978.04 15905.97 f 622.14 196.19 0.00 7668.56 11978.04 12076.80 42L 10A 7668.56 12076.80 -11978.04 42R 10A 622.14 0.00 -196.19 43L 10B 0.00 622.14 -196.19 ~7609.44 3201.20 -26144.33 43R 108 0.00 196.19 622.14 -7609.44 26144.33 3201.20 ' 44 0.00 196.19 622.14 -7609.44 48827.14 -3951.87 45 0.00 196.19 622.14 -7609.44 71510.08 -11104.98 5 i 46 0.00 196.19 622.14 -7609.44 94193.00 -18250.08 47 H15 0.00 196.19 622.14 -7609.44 116876.01 -25411.21 48 0.00 196.19 622.14 -7609.44 136809.91 -31697.40 49 0.00 196.19 622.14 -7609.44 156743.88 -37983.63 l 50 0.00 196.19 622.14 -7609.44 176677.88 -44269.84
-7609.44 196611.84 51 0.00 0.00 196.19 196.19 622.14 622.14 -7609.44 216545.94
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.I, ARKANSAS POWER & LIGHT SHUTDOWN COOLING LINE - FRE1.1HINARY ANALYSIS RESULTS 340 DEG F DELTA T STRATIFIED FLOW CASE i
LOAD CASE NO. 4 (SFL1). FORCES AND HOMENTS IN 1.OCAL COORDINATES RUN SOP DCP AX1AL Y Z XX YY ZZ GROUP HMB NAME FORCE FORCE FORCE HOMENT HOMENT HOMENT (LB) (LB) (LB) {LJ.IN) (LE.IN) (LB.IN) l PRN1 1 1 9126.64 0.00 439.72 113314.08 2008512.00 -1519044,00 2 9126.64 0.00 439.72 113314.08 2104652.75 -1519044.00 3 9126.64 0.00 439.72 113314.08 2120793.75 -1519044.00 4 9126.64 0.00 439.72 113314.08 2136934.25 -1519044.00 SL 2A 9126.64 0.00 439.72 113314.08 2153075.50 -1519044.00 /$TAhnt A6Md g62-SR 2A 9126.64 -310.93 -310.93 113314.10 -448327.81 2596580.50 6L 2B 310.93 9126.64 -310.93 455529.47 106112.43' 2392395.75 6R 2B 310.93 -9126.64 310.93 438837.25 -102224.10 -2304730.00 7 3 310.93 -9126.64 310.93 438837.25 -99787.63 -2233213.50 BL 4A 310.93 -9126.64 310.93 438837.25 -97318.30 -2160734.25 BR 4A 310.93 -310.93 -9126.64 438837.28 2160734.50 -97318.38 9L 4B 439.72 0.00 -9126.64 -1157920.13 1694179.25 -94444.66 9R dB 439.72 -9126.64 0.00 -1157920.25 -94444.66 -1694179.25 10 439.72 -9126.64 0.00 -1157920.38 -94444.67 -1480342.50 11L SA 439.72 -9126.64 0.00 -1157920.25 -94444.66 -1266504.88 11R SA 439.72 0.00 9126.64 -1157920.13 -1266505.00 94444.66 12L SB 310.93 310.93 9126.64 17136.08 -1570332.00 91570.95 12R 5B 310.93 -9126.64 310.93 17136.07 -91570.95 -1570332.00 13L H2 310.93 -9126.64 310.93 17136.07 -80563.98 -1247249.63 13R H2 310.93 -9691.00 310.93 17136.07 -80563.98 -1247249.63 14 6A 310.93 -9691.00 310.93 17136.07 -79242.50 -1206062.38 15L 6B 9691.00 310.93 310.93 74617.33 21761.24 -1066532.25 ISR 6B 9691.00 0.00 439.72 72478.49 747481.56 -717588.63 9691.00 0.00 439.72 72478.49 761096,13 -717588.63 16 H3 17 H4 9691.00 0.00 439.72 72478.49 764726.75 -717588.63 ' 18L 7A 9691.00 0.00 439.72 72478.49 769718.01 -717508.63 18R 7A S591.00 -310.93 -310.93 72478.48 -36861.54 1051685.13 19L 78 310.93 9691.00 -310.93 43600.43 65739,59 848389.25 19R 78 310.93 -9691.00 310.93 42396.02 -63923.61 -824953.25 20L HI 310.93 -9691.00 310.93 42396.02 -59186.11 -677297.06 20R HI 310.93 -9677.31 310.93 42396.02 -59186.11 -677297.06 21 VA1A 310.93 -9677.31 310.93 42396.02 -50761.17 -415083.19 3{ 22 310.93 -9677.31 310.93 42396.02 -44200.45 -211139.47 'fg L. 23 VAls 310.93 -9677.31 310.93 42396.02 -37655.71 -7194.95 24L BA 310.93 -9677.31 310.93 42396.02 -37639.63 -6694.64 -
* (7 24R BA 310.93 310.93 9677.31 42396.01 -6694.66 37639.63 S 25L OB 0.00 439.72
-9677.31 9677.31 439.72
-25021.51
-25021.51 169454.95
-31086.90 31086.90 169454.94
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s 25R BB 0.00 8 81 ! 26 H5 0.00 -9677.31 439.72 -25021.51 -30071.29 191806.41
- l 27L H9A 0.00 -9677,31 439.72. -25021.51 -27423.50 250076.14
-9677.24 439.72 -25021.51 -27423.58 250076.14 (T 27R H9A 0.00 i CN d' s P O
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r l-ARKANSAS POWER & LIGHT SHt'TDOWN COOLING LINE - PRELIMINARY ANALYSIS RESULTS 340 DEG F DELTA T STRATIFIED FLOW CASE i LOAD CASE Ho. 4 (SFL1), FORCES AND HOMENTS IN LOCAL COORDINATES (COffrD. ) SOP DCP AXIAL Y Z XX YY ZZ RUN MOMENT GROUP MMB NAME FORCE FORCE FORCE HOMENT MOMENT (LB) (LB) (LB) (LB.IN) (LB.IN) (LB.IN) FRN1 (CONTD.) -25021.51 -20804.32 395749.47 20L H9 0.00 -9677.24 439.72 28R H9 0.00 468466.31 439.72 -25021.51 -20804.32 395749.47 29L H9B 0.00 468466.31 439.72 -25021.51 -14185.06 -6656163.50 29R H9B 0.00 -68642.08 439.72 -25021.51 -14185.06 -6656163.50 30 VA2A 0.00 -68642.08 439.72 -25021.51 377.31 -4382943.00 31 0.00 -68642.08 439.72 -25021.51 9644.25 -2936350.50 32 VA2B 0.00 ~68642.08 439.72 -25021.51 18911.23 -1489752.50 33L H17 0.00 -68642.08 439.72 -25021.51 29943.33 232384.92 33R H17 0.00 758.94 -131.12 -25021.51 29943.33 232384.92 34L 9A 0.00 758.94 -131.12 -25021.51 27100.14 216390.75 34R 9A 0.00 131.12 758.94 -25021.49 -216390.75 27180.14 35L 9B -131.12 0.00 758.94 200396.56 -9027.32 24416.96 35R 98 -131.12 758.94 0.00 200396.55 24416.96 9027.32 36 H21 -131.12 758.94 0.00 200396.55 24416.96 6742.44 37 -131.12 758.94 0.00 200396.55 24416.96 -19762.08 30 -131.12 758.94 0.00 200396.55 24416.96 -46266.72 39 -131.12 758.94 0.00 200396.55 24416.96 -72771.37 1 40 -131.12 758.94 0.00 200396.55 24416.96 -99276.00 41 H16 -131.12 758.94 0.00 200396.55 24416.96 -125780.80 42L 10A -131.12 758.94 0.00 200396.55 24416.96 -139490.09 42R 10A -131.12 0.00 -758.94 200396.56 -139490.09 -24416.96 43L 10B 0.00 -131.12 -758.94 155484.30 184402.38 -21653.77 43R 10B 0.00 758.94 -131.12 155484.30 21653.77 184402.38 44 0.00 758.94 -131.12 155484.30 17229.40 158792.75 ' 45 0.00 758.94 -131.12 155484.30 12805.01 133183.02 46 0.00 758.94 -131.12 155484,20 8380.62 107573.26 47 H15 0.00 758.94 -131.12 155484.30 3956.22 81963.42 40 0.00 750.94 -131.12 155404.30 68.04 59457.43 49 0.00 758.94 -131.12 1554a4.30 -3820,16 36951.32 50 0.00 758.94 -131.12 155484.30 -7708.36 14445.21 51 0.00 758.94 -131.12 155404.30 -11596.55 -8060.88 g 52 11 0.00 758.94 -131.12 155484.30 -15484.77 -30567.09 g* (.o b D h D. i b t M 0 0
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l 1 l REFERDJCES
- 1) APfL Dwg 2CCA-25-1, Rev 9, Large Pipe Isometric - Safety Injection System
- 2) Bechtel Dwg 20CB-5-4, Rev 06, Isometric - Containment Build-ing Safety Injection
- 3) Entergy Operatiens Dwg 2CCA-25-H9, Rev 09, Hanger Detail, SI System
- 4) Bechtel Dwg H-20-517, Rev 08, Pipe Support, Reactor Building SI
- 5) ASME Boiler and Pressure Vesel Code, Section III, Class 1 Components, Division 1, Subsection NB, 1986 Division, and Appendicies.
- 6) ABB Letter ME-93-067, G.A. Pierfedeici to B.T. Lubin.
Preliminary Stress Analysis of ANO2 Shutdown Cooling Line with Thermal Stratification, dated 6/8/93.
- 7) ABB Letter A-MECH-93-012, B.T. Lubin to R. Lane, Return to Power of Arkansas Nuclear One - Unit 2 from Outage 2P-93-1, dated 5/11/93
- 8) Bechtel Dwg 2CCA-25-2, Rev 6, Large Pipe Isometric - Safety Injection System
~
ABB Combustion Engineering Nuclear Power
,_ , neemma. m
. /?)}$(~ A*0 *(* )$Y & l Calculation Number Rev. l l l l 23 l Page Number
. 9) Calc MISC-ME-C-057, Rev 00, Stress Analysis of Surge Lines for CEOG Task 587, Including Effects of Stratified Flow, dated 8/7/89
- 10) Calc MISC-ME-C-057, Rev 02, dated 5/13/91
- 11) Eeer, F.P. and Johnsten, E.R., " Vector Mechanics for Engineers -
Statics and Dynamics', McGraw-Hill Book Co., 1962
- 12) Calc K-ME-C-021, Rev 01, Surge Line Routing, Support System, and Stress Analysis for YGN 3 & 4, Including Effect of Stratified
~
Flow, dated 12/9/88
- 13) Tube Turns, " Pipe Fitters Manual', Chemtron Corp., 1977 L
COMPUTER CCDE REFEREJCES NAME PF ID/ VERSION COMPUTER USED Superpipe spipe / Vers 22E HP9000/400T (Desktop) _2_ ABB Combustion Engineering Nuclear Power
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= /37, / 2.s z
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- gn.'ll t i ARKANSAS POWER & LIGHT SHUTDOWN COOLING LINE COORDINATES CONTROL POINT COORDINATES, AS COMPUTED AND STORED
. i t
RUN POINT POINT GLOBAL COORDINATES ;
- NAME NAME TYPE I . Y Z (IN) (IN) (IN) l PRN1 l 1 0.000 0.000- 0.000
- 2A TNP 0.000 -133.125 0.000 2* TIP 0.000 -154.125 0.000 2B TNP 14.849 -154.125 14.849 i 3 20.064 -154.125 20.064 ;
4A TNP 25.349 -154.125 25.349 l 4* TIP 31.500 -154.125 31.500 ! 4B TNP 31.500 -154.125 40.198 SA TNP 31.500 -154.125 84.302 [ 5* TIP 31.500 -154.125 93.000 ) 5B TNP 37.651 -154.125 95.151 ; H2 61.210 -154.125 122.710 125.538 6A TNP 64.038 -154.125 ! 6* TIP 73.938 -154.125 135.438 i ) 6B TNP 73.938 -168.125 135.438 , i H3 73.938 -198.125 135.438 ! H4 73.938 -206.125 135.'.38 l 7A TNP 73.938 -217.125 135.438 7* TIP 73.938 -238,125 135.438 ,
- 7B TNP 88.787 -238.125 150.287 i H1 99.522 -238.125 161.022 l VA1A 118.614 -238.125 180.114 !
VA1B 148.313 -238.125 209.813
- 8A TNP 148.349 -238.125 209.849 ;
8* TIP 154.500 -238.125 216.000 ! 8B TNP 163.198 -238.125 216.000 l H5 165.500 -238.125 216.000 , H9A 171.500 -238.125 216.000 . H9 BRP 186.500 -238.125 216.000 ! H9B 201.500 -238.125 216.000 ! VA2A 234.500 -238.125 216.000 VA2B 276.500 -238.125 216.000 H17 301.500 -238.125 216.000 l
- 9A TNP 322.500 -238.125 216.000 ;
i 9* TIP 343.500 -238.125 216.000 ; 9B TNP 343.500 -238.125 237.000 {
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, RUN POINT POINT GLOBAL COORDINATES NAME NAME TYPE X Y Z (IN) (IN) (IN) l l
PRN1 (CCNTD.) H21 343.500 -238.125 240.000 L H16 343.500 -238.125 414.000 i 10A TNP 343.500 -238.125 432.000 10* TIP 343.500 -238.125 453.000 t 10B TWP 364.500 -238.125 453.000 l HIS 499.000 -238.125 453.000 11 646.750 -238.125 453.000 l PRN2 l 7 H9 BRP 186.500 -238.125 216.000 t
- H7 186.500 -165.125 216.000 l j 19A TNP 186.500 -154.125 216.000 l
- 19* TIP 186.500 -142.125 216.000 ;
19B TNP 174.500 -142.125 216.000 20 170.500 -142.125 216.000 ' a J l l a 1 1 . A e i
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1 J 1 f APJ.ANSAS POWER & LIGHT SHUTDOWN COOLING LIIE - STRATIFIED FLOW < ark.2f.rev> ADD 8-INCH LINE REPRESENTATION i USE TEMPS IN BTL LETTER TO ANO 5/11/93
- AND EXACT SPRING STIFFNESSES; H9B BILIN; H9A INACTIVE ,
N FL gg Ac7w . c C EXI'd STAT COMB DESC C a C USE ! EXT CARD FOR DATA CHECK ONLY C 3 CCHEK DESC E-80 C C CARD B4- UNITS (BLANK FOR DEFAULT) ; I C l C SECT. C1% TITLE OF GEOMETRY SET l ARK 2 0 E-80 XPRN ARK SHUTDOWN COOLING LINE
! C C SECT. DIA% PIPE RUN NAME & TITLE PRN1 CLS1 ARK SAFETY INJ 14-INCH LINE 1 DIR 0.0
. 2* TIP 21.0 CFF -154.125 1 3 STL 28.375 2* 4*
4* TIP 21 OFF 31.5 +31.5 2* 5* TIP 21 OFF 61.5 4* H2 STL 18.00 6* 5* 6* TIP 14 OFF 42.4375 +42.4375 5* H3 OFF -44. 6* H4 OFF -52. 6* 7* TIP 21 OFF -84. 6* H1 STL 27. VA1A 7* VA1A STL 42. VA1B 7* VA1B STL 8.75 8* 7* 8* TIP 21 OFF 80.5625 80.5625 7* H5 STL 11. 8* 9* H9A STL 17. 8* 9' H9 BRP STL 32. 8* 9* H9B STL 47. 8* 9* VA2A STL 33. H9B 9* VA2B STL 42. VA2A 9' H17 OFF 25. VA2B 9* TIP 21 OFF 189. 8* H21 OFF 24.0 9* H16 CFF 198.0 9'
NI 6 C- M 6 " b' (6N Q O. STRP BS140 BELB 85140 P- A 3 BELL 145140 MASS 13.333 1.724 1.288 60. BELB 14S2C MASS 13.333 1.724 1.288 60. VALV VALVE 20.0 4.0
*FGBW FSBW C i C CARDS F% MATERIAL FROPERTIES ,
C C SA403 WP316
*SA312 TP316 t STEP STRP 145140 SA312 TP316 2A i FGBW FGBW FGBW ,
. BELB BELB 145140 SA312 TP316 2B FGBW FGBW STRP STRP 4A FGBW FGBW BELB BELB 14S140 4B l FGBW FGBW STEP STEP SA FGBW FGBW BELB BELB SB j FGBW FGBW STRP STRP 6A I FGBW FGBW BELB BELB 145140 6B I FGBW FGBW STRP STRP 7A
~
FGBW FGBW l
,, BELB BELB 7B FGBW FGBW l STEP STRP VA1A !
FGBW FGBW VALI VALV VALVE SA312 TP316 VA1B i STRP STRP BA i FGBW FGBW l BELB BELL 145140 BB
, FGBW FGBW STRP STRP H9A HAN9 VALV M9B l
STRP STRP VA2A VAL 2 VALV VA2B STRP STRP 14520 9A l FGBW FGBW BELB BELB 14520 9B l STRP STEP 10A FGBW FGBW BELB BELB 10B FGBW FGBW
*STRP STEP 11 STRP STRP 85140 SA312 TP316 19A PELB BELB 85140 SA312 TP316 19B
'STRP STRP 20 C
C SECT. Hi LUMPED WEIGHTS e Misc-- MG- C- (64 C U OI C SECT, It SUPPORTS ! HOTL 1 AUCH P- A 4 l HNG1 H1 HA!JG 1600.0 Y ~ HNG2 H2 HANG 1200.0 Y HNG9 H9 HANG Y HN9A H9A HANG 1.0 Y I C l I
. C H9B BILINEARIZED '
HN9B H9B HANG Y l H17Y H17 HANG Y H172 H17 HANG ' ENDS 11 ANCH U HNG7 H7 HANG X
*AN20 20 ANCH f
. C C CARD J% OUTPUT POINTS !
SOPS 36.
*DOPS 36.
C C SECT. Li STATIC ANALYSES EXPO THPR NOPR ZERO-LCAD AUTE FFPR 0.
*RUNL 70. 70. ,
EXPN THEM NOPR THERMAL - 500 DEG CONST TEMP AUTE ! FFPR 2250. RUNL PEN 1 500. 70. I 11 SUPT HOTL .7706 +.3353 0.
*SUPR HOTL +.00106 i . EXP1 THRM PRIN THERM EXPN-STRATIFIED CASE AUTE FFPR 2250.
RUNL PRN1 560. 70. 1 2B i RUNL PRN1 385. 70. 2B 6B RUNL PRN1 80. 70. 6B 7B 1 i RUNL PFlil 80. 70. 7B 11 ' SUPT HOTL .8835 +.3844 0.
, *SUPR HOTL +.00122 STRI THRM PRIN STRATIFIED FLOW INPUT FFPR 2250.
TEME PRN1 560. 70. 1 2B TEME PPlJ1 385. 70. 2B 6B f TEME PRN1 80. 70. 6B 7B TEME PRN1 80. 70. 7B 11 C , C THESE ARE SCALED BY 350/340 SUPT HUG 2 +2.6810 i SUPT HNG1 +4.9221 SUPT HNG9 +8.8793 i SUPT HN9A +8.5629 ! C
~
C DEL-Y M9B = 9.1955 + 1/16* SUPT HN9B +9.2580 SUPT H17Y +12.5055 SUPT H172 +3.2441 SUPT ENDS +1.7710 +27.7338 +3.2441 j i I
Mt6C- Ll,E - C. I6 4 E 01 i SUPR ENDS .03862 SUPT HNG7+0.2321 0.0 +.02108 p- 4 [ l SUPT AN20-0.2532 +8.5451 -0.4632
*SUPR AN20 .03862 0.0 +.02108
'GRAV GRAV !!OPR DUMMY WEIGHT LOADING ,
TDG 650. 650. MASS :.0 1.0 1.0 4 FFPR 2250.
*GRAV C
C SECT. PA- CCMB OPTICH tPPINT CCMB'S FOR SUPPORT LOADS-L DISPS)
~
C ,
~
C ALSO DEFINE LCAD CASES FOR DESIGN CHECR I GRAI GRAV NCPR GPAV 1.O ! THMO EXPN NCPR EXPO 1.0
. TH!C EXPN PPliX EXPN 1.0
*SFL1 EXPN PRNX DSUM EXP1 1.0 STR1 1.0 C .
C SECT. Tt CLASS 1 DESIGN CHECKING C C SECT. T1% CONTROL CARD
*CLIA CETL OLDC NEWP NEWT C
C SECT. TS% PRESSURE DISTRIEUTIONS j PRDO PRESSURE DISTRIBUTION FOR PD/2T (ZERO LOAD) 0. PRDD PRESSURE DISTRIEUTION FOR PD/2T (DESIGN PRESSURE) , 2500. PRD1 PRESSURE DISTRIBUTICN FCR 440-120 CASE ' 410. ;
*PRD2 PRESSURE DISTRIBUTION FOR 653 CASES 2250.
C
. C SECT. T6% TEMPEPATURE DISTRIBUTIONS
- DiD1 TEMPERATURE DISTRIEUTION FOR SM DETERMINATICN 700.
c C CECT. T7% CEFINE LCAD SETS FOR EQ 9, DESIGN CCNDITICN *
- DES DCCN IRDD TMD1 GRAI DES PRES + WEIGHT C
C SECT. TSA% CCNTRCL CARD FCR NB3653 CONDITIONS O .1 2.4 100 3.0 2.4 0.1 !0B3 653
- C C T9- LCAD SETS FCR NE3653 CCUDITICNS.
C , C USED TO DETEPl4INE STRESS PANGES. ZERO 1000 PRDO TMD1 THMO ZERO-LCAD SFL_ 1000 PRD2 MD1 SFL1 340 DELTA T CASE
*T500 1000 PRD2 TMD1 THMN 500 EEG CASE
_4
y I i 1 1 3 s
. i t
t t i a r i t APPENDIX C : 1 to , t i 1 e
; 1 l
ABB-CE Report No. A-MECH-ER-009. Rev.00 i
-l I
?
l l e s i i JB W il 9 'f 1 i
ABB ASEA BROWN BCVERI Calculation !4 Pages Contract m 2.1% - t Appendix 412-Pages , Microfiche NA Calculation Number A- MM - f_ A LC - 0 2 3 Revision 00 Title Ir&* munwt, Sets and Y1tave At\bss S KamtL %+. Ces h m m M J J
- JND2 Snutdown (onhna Line 2CC A - 2 5- P4" J
Author I- 6 b ne h Date S I 93
\ \ ' '
Calculation contains safety related design information: Yes <- No VERIFICATION STATUS: COMPLETE The Safe >Rcwexf der #;n H:rmallra,arilshed in tils docuent has toEritodell b t.s carryt by means at
- v Deson Reviewue:rg CheckAstfs) O cfN10L Anemaw Anayess . Copy a3 acted.
Veri 6canonToseng TestReportN1 - faler+ C. VL es.kr /C[.-s2 C< W- l ~ 8 9y in&pondard Reviewor NanWE* mat.reOSW j g lj l / ' A / p Approved by 3 ~' M 'n ;$ceror, r - Ef,jvcffhu.te,3 Levia r) Date
; b. >
Distribution 3 2"un ( 2) G ]s rSt kici Summary
Purpose:
-i no prhrm a pl+n aty &t Lt end hf,3 sit ugl s,.c y > r' he fHDz SnJQn Gol,n U% :n :w sik a b.m f precusl y uud$ ud. M,ft1 t,ridi:ms.
Method and Results of Review:
Tha :.ach+,6 4s terl e) 5 3 % u.%4 4 kf %ew anL sOdes, ave 9 a.ble, h Ss Jeme on ChecU.s F H o. 2 :s % &4 Asw.ua %ahr
4.1 Assumptions
- 1. The Code of Record for the SDC line is the ASME Boiler and Pressure Vessel Code, 1971 Edition through the Summer 1972 Addenda (per Reference 5). Initial evaluations of the effects of the stratified conditions indicated that usage factor results may be unacceptable if it were calculated in accordance with the Code of Record. The primary reason for this unacceptability is due to the fact that the Code of Record categorizes the AT, (linear gradient) stress as a Secondary stress. One potential - and significant - effect of tnis categorization is an increased Primary-plus-Secondary Stress Intensity Range (as calculated in Equation
- 10) such that a [ greater) K, multiplier would be used in calculating the alternating stress (per Equation 14); a higher alternating stress results in a higher fatigue usage ,
factor. The 1986 ASME Code considers the same AT, stress as a Peak stress only. ABB Combustion Engineering Nuclear Power
1 l A BD R ) JM p p ME A BAOWN BOVERI A - M ECH - CA LC.- 02 3 00 i Calculation Number Rev.
. l l
O $0 Page Number The use of the 1986 Code for this evaluation is justified in that this evaluation is being performed to assess the structural integrity of the SDC line for near-future operations in consideration of the previously unanalyzed
~
~
stratified conditions. Use of the 1986 Code, in spite of l the difference in categorization of the aT, stress term, still results in conservative values. Additionally, there is precedent in the use of a later Code (and non-Code of Record) for fatigue analysis of thermally stratified lines, as permitted in NRC Bulletin 88-11 (for Pressurizer Surge Line Thermal Stratification; dated December 20, 1988). ,
- 2. The radial gradient (through-wall or local) stresses for the stratified conditions are accounted for by incorporating a ratio-ed stress based upon analysis of stratified conditions 1 in a pressurizer surge line. Reference 6 accounted for >
stratified conditions for a 320*F top-to-bottom AT for a 12-inch, Schedule 160 (stainless steel) line, with a local 4 rtress value of 15.3 ksi. When multiplied by a ratio of 350/320 (to account for a larger SDC line AT), the stress value becomes 16.7 ksi. Though the surge line and SDC line pipe sizes and schedules differ, and though conditions internal to the pipes may differ, it is assumed that the 1 difference in resulting steady state local stresses is minimal such that use of the surge line analysis is acceptable. 1 l For calculation of the Peak Stress Intensity Ranges (ASME , Code Equation 11) in Section 4.2, the local stress value is assumed to be additive, without consideration of the actual
, algebraic sign. This assumption is conservative since the the resulting Peak Stress Intensity Range is maximized.
- 3. An Sm value for the elbow material (SA-403, WP-316; see Figure 1) was not directly obtainable from the appropriate table of the 1986 Code (Table I-1.2). Note (4) of this i
Table specifies that the Sm value for SA-403, WP-316, is to be "the same as those assigned to the material from which the [ elbow) is made." Consequently, an S, value was obtained by first correlating the S, values used for Data i Point 10 in Reference 4 (Page 1128) to a material (s) whose L 4 S, values are given in the Code of Record. (For example, the S, value used for Load Set Pair 0--I was 57600 psi /3 = 4 19.2 ksi (T=400'F). Also, the S mvalue used for Load Set
, , Pair 0--15 was 50700 psi /3 = 16.9 ksi (T=611*F).) The Load :
Set S, values correspond to a set of high alloy steels given on Page 396 of Table I-1.2 in the Code of Record. A
, corresponding S, value for the same material (s) was then ABB Combustion Engineering Nuclear Power
A B ER MND ASEA BROWN BoVERI A - ftECl4 - CALL- 02 3 C)0 Calculation Number Rev. 50 k Page Number selected from Table I-1.2 of the 1986 Code. The temperature at which the Sm value was chosen is 400*F.
. This temperature is the approximate average temperature for the stratified conditions; it is higher than the average temperature experienced by the line for either of the design basis Load Cases 14 or 17 and, therefore, results in a lower, more conservative S, value.
- 4. Load Combination No.1 is used in this calculation to designate the Load Set consisting of Load Case No. 17 -
Cooldown with Initiation of Shutdown Cooling (Ref. 4) and the 350*F Thermal Stratification conditions. In order to account for fatigue effects of this Load Combination, it will be assumed that this Load Combination has occurred the same number of times that the plant has had a heatup-cooldown cycle. Review of the information provided in the table in Appendix A entitled " Unit 2 RCS Heatup/Cooldown" indicates that ANO2 has had a heatup/cooldown cycle a total of 68 times, up to and including 2P93-1 (5/1/93).
- 5. Load Combination No.2 is used in this calculation to designate the Load Set consisting of Load Case No. 14 - S.I.
Check Valve Test transient (Ref. 4) and the 350'F Thermal Stratification conditions. In order to account for fatigue effects of this Load Combination, it will be assumed that this Load Combination has occurred the same number of times that the plant has had either a Power Reduction, a Reactor Trip, or Loss of Reactor Coolant Flow. Since the number of Power Reductions is available from 2R8 (see the table "ANO2 Power Transient History 2R8 to Present" in Appendix A), the number of Power Reductions prior to 2R8 will be "backfitted" by determining the rate at which Power Reductions have occurred since 2R8 and tripling that rate to estimate how many occurred prior to 2R8 back to Startup (see Section 4.3).
4.2 Details
Stress Considerations Evaluation of the stresses from the design basis analysis (Ref. 4) indicates that the primary contributors to the fatigue usage factor for the elbow are Load Cases 17 and 14
. (Cooldown - with Shutdown Cooling Initiation - and the S.I.
Check Valve Test, respectively). Consequently, each of these will be evaluated in a Load Combination with the Thermal Stratification condition. ABB Combustion Engineering Nuclear Power
ABB ASEA BROWN BOVERI 4- ME(4-C A LC- 02 3 00 Calculation Number Rev. 2 d 19 Page Number
. LDad (om b\on % n No.I:
Lod Gsc W. G ark 3504 ~ThernwL Shb Auhn h5mE cD3E'-REE 3] () tG - 3& f3,) h o^ry N ^arn6q arr In4vut -7f by , En i En = d i 4 yM i - G Es x j 4L a,7 j 535s f" 3 pbphn 10 C,= ;.22 -hi 2 ((cwphr % tio.s.) Jz: 2.6 l Q= 1. 0 0
; ?, = ; 385 - n 3E l p u = MD 96 1
3B6 p. -7d. 4,p 11 123Ep - Ref 4 , p Z I a Do= En - h_pc 1 er 14" Ica 170
- Rtf } , $r Irmenc trM j
+= . t, zs ,o n i
I= 1027in'4
, 4.T - d5T =0 i A sm. 19.3 -Tef 3 , ~Tdl* L l2. [O900' N (W 4* 4t* 3)
ABB Combustion Engineering Nuclear Power I f arm
- 0090257 B tRev 7 90)
~ ABB ASEA BROWN BOVERi A-MFZH - C A LC.- 023 AD Calculation Number Rev. h sf Ih Page Number All is ojalilt4, x 6/ lows :
=
f thf 1 ,3 007 -! #- e inc7 4-Ib - 2D32 4 !b. 4r exw W) ~ . _ , , , .
-9,1in-suff < f) 4- f.y' s .:4t n- cp 35O'F Tkres %%% :n, b $
'l2.8m kyr 57f w hip 2475 n- 4 O%L +o hf + g 2. g Sy -21M vr<en trbon UtMES %.- /(b. Mc l 12A w>5 235, ,
1 M; = y /RI ma'
- me '
23 87 in - bp
) En e 12..G
- 42.6 + 0 = Ss. I ku 6 3Sm = 57.9 ai L
ABB Combustion Engineering Nuclear Power i j Form a 0090257 B (Rev 7.9 01 '
l ABB ASEA BROWN BOVERI f.MRl4 -Ctt Lc- 023 DD Calculation Number Rev. l0 .J I9 Page Number
- 5) !!3 . % 5 3. 2 ?uk 3rc:in% d4 % ee, ip:
S -(,C,P',b"
- K: C. 3y&.@-9 )AT,l K, +&K3 C, Eg ( l4TcqTi, }
+
G \Ar.I Et uka U (,= w K= 3 !. D - Rel' 1 (Gh. Lf Mo (,) 9- 0.3 -R6 4 , p. H D5' E- 23.3 e 10' ps t -M 3 p- " BMx10"<g.,p W bcad %th.0= .~ D3 ' ? - kV +,p lIOS E -ir Lu.L 0x ))o.47 - l2 0* F
- 3. t. c.t.b d 3rdko+ kosak-usU ) stressa r's es 2509 3tra ibnl fiul w W M a n sol k uavnttL ?:r bus .tuin3 +t envin-A a seu idut 11, llo.1 ksl (ut Asia 4mn 2. ) .
~
> Q = YZ. (c. + 42. i - 3 4. (, - 40.6 u D + \ {, .1 = i+ M ksL i
ABB Combustion Engineering Nuclear Power l Form a 0090257-B tRev 7 90)
A4B ASEA BROWN BOVERI A NEC4-(_ALC-o7.3 00 Calculation Number Rev. Il Dh !h Page Number
. :)(15-3LS3.(, .5,mpiJeL % bc- Tlashc h ts urd.nud, A ndas,s
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- 3) G - 365 3.7 12rnul Shur 1.dchtt h,ahm Io a hasind hr L u L Cc m 6,o n h an 5nu. ti a l, czmkhkn dk bT, ru e a nt+ ttiwret ,
u 0,nso w +\ , yr N 3 - 165 3.3 , 4 t. Alkrr. desa :wess eer su a a 3 ; 4r d Cmbedu No.1 is crioattel u t ollos3. Seu = SL
, A 2.
2 4
"2. G kst
~
ABB Combustion Engineering Nuclear Power Form # 009025,7 9 (P,ev 7 90:
ABB ASE A BROWN BOVE Al A- M ECH- CA LC- 0 23 00 Calculation Number Rev. 12 J 19 Page Number
. a s L Grc hi. v :1. bio. 2 LwL asc 9 o . F+ .ad 350 F 7,errnd. ht 6ahon A) '16- 16 53. !
Ali vrmr af. me. ;ted L: $r aid (crne:cdion %. i u czp +-
% = lI 2235 - 14 3E , =
60 0 pst 2235jrt -Tef 4,p.2 Ib35 yst -
- Utt2Ndwel r1went)
Low far luderTr;p Ipd J CAe fj o. A Il 4 41 6 [h
~
i 11h l C. .b - 13 41f Q. L5 10694 EMb Od? 4 ,pIDBG) l
=
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% =
2A 04 ,n- 4
.) En = +f e II.7 e D = Is.6 kri 6 57.9 bt ABB Combustion Engineering Nuclear Power I
rarm
- 0090257 81 Aev 7 90) '
l l A R BD l 7%B59 ASE A BAOWN BOVERI 4- MECW - CA LC.- 02 3 00 l1 Calculation Number Rev. 13 ot !9 i Page Number
. 5) (16-3653.2.
For Ltd Gu li+ 6T, = - iz B"F - ?d. 4 , p. i o r it Art = -32. F
-S p
= el
- n? - 21. 6 - 0 + I D.4
- I6.7 = 105. 2 kri ,
i l M kt = E: 2.
- 52. 6 ksi i i
i f 1 i ABB Cornbustion Engineering Nuclear Power Form a 0090257 B 4Rev 7 90;
7% B B A',E A f?POV N BOVEm A - M ECH - CALC- 02 3 00 Calculation Number Re v.
- + Of !@
Page Number
4.3 Details
Fatigue Considerations Consideration of fatigue effects is accomplished by the
~
evaluation of cumulative fatigue at the same weld which was evaluated for stresses (in the previous section). Emphasis for the fatigue effects will be upon determining the , increase in the fatigue usage factor to-date and for the ' near future. It is noted that the limiting location does not have the highest usage factor for the entire line. However, the increase in the fatigue usage factor for this location for consideration of thermally stratified conditions effectively bounds all locations in the line. 9 ABB Combustion Engineering Nuclear Power
ABB ASE A BROWN BOVERI j,(*1Els4- (JiLC - 02 3 00 , Calculation Number R ev. ! f
! $ ,,f' !h Page Number b )CT(l AIkl ATICtl DF 'RUbi C IJT OUVELENLES Euf 'D J brsnNATID6) VID, E Mr %etwas >
Ikkr J br Rehd+nr a & +ao it) p a sin a 2R B I (, Gh.m kJ n~&r <f br %Lons ,a 4 brtm (i3 ) prs l > ar = :RB-4w . x3 < 13 = h 1 t : I
% pr gar saa 2R6 i
i t lde 7n or +, 2R8 bgfe4 e d anx 2R9 { i t i Omkr a f prt y t,r b ZFS
%ev TF;p .wL Less C hr,r % f ae+ Fows.
%Gr = { 103 + l} + l+ =
ICS I , t Whee Tr p wa hyed %L,r Th( a r WL h zcz w TRAusicuT5 Tb-bArc 4% i ABB Combustion Engineenng Nuclear Power form a 0090257 B ;Rev 7 906
=. .
ABB ASEA BAOWN BOVERI A-14EClf-U1 LC- 02 3 DD j Calculation Number Rev. ^ l& S lb Page Number
- 3) To .htL Y&aut Jstart a
Yae for ld U.k%n .
%u na t1 u-L:d Cwb % I T3.hksi % 90 [* b C\ 2.
O.oz9
} _:,4 u b. w : 57.t.kn u,200 '3 !, .or7 -
l by 3tns aeqe 6*r (&f4,p1128) OM ' 4 ! ~Tcfd_ tt 0.073 i 1 C) Nur Ydwt Yd y et. Ulyr Ynbr Celerb hen 4
-k lha Tvtvr e v< Vuoc Fathr udI be r4 & dd. m ht-unsphtc --n t+ see Nib)d i e J
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i 000 (:rt Am u l ) .:L 4 h r u L J cwer
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0.Cb9 Ltd lemb. Ab. 2 52./, bi 24,2 00 436 ,Of$ - ! kss?cas Luua inne DW T+d. O C. I 13 1
. h = 0.f \ 3 - 0. 013 = 0.040 1 i t
1 ABB Comoustion Engineering Nuclear Power ] Form
- 0090257 B IRev 7 903 h
)
l A BR R MBBBB ASE A BACWN BOVE Ri A ft ECH - CALC- 02 3 00 Calculation Number Rev. I T Ob h k Page Number 1 - 1 . According to the 1986 ASME Code, then, the increase in the fatigue usage factor for the elbow is 0.029 to-date and l 0.069 for the near future (10 additional Heatup-Cooldown
- cycles and 1000 additional Power Reductions, including Trips, etc.). For the location on the line with the highest #
usage factor - 0.398 at Data Point 63 (per the Code of t Record and Reference 4) - the overall usage factor for the f near future becomes 0.398 + 0.069 = 0.467. It is noted that this value is based upon the occurrence of all design basis transients as well as the additional Load Combinations. 1 I i
. I A
P i 4 l 2 i P ABB Combustion Engineering Nuclear Power
A RD ! Mpp ASE A BROWN BOVERf A ttE<H -CALC- 02 3 00 Calculation Number Rev. l I$ $ l$ Page Number ( i
5.0 REFERENCES
, i
- 1. Arkansas Nuclear One Contract Order No. 103, under Contract F.o . 1s-1007.
- 2. ABB CE Calculation No. MISC-ME-C-164, Rev. 01, July 1993.
- 3. The ASME Boiler and Pressure Vessel Code, Section III, 1986 Edition, no Addenda. '
- 4. ANO2 Report No. 85-E-0055-21, Rev 00 (including Class 1 Stress Analysis of the Shutdown Cooling Line).
- 5. Specification No. 6600-M-2200, Rev. 09, " Design Specification for ASME Section III Nuclear Piping for Arkansas Nuclear One - Unit 2."
- 6. ABB CE Calculation No. MISC-ME-C-057, Rev 02, May 1991.
- 7. Crane Co. Publication No. VC-1900A4, " Engineering Data Catalog," 1976.
ABB Combustion Engineering Nuclear Power
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mY 06 '93 4:25 TEmHON SUPPNT ' ' v. Qf wasse z _
"ME153E .XLS 2e._
h- : c AND-2 Power Transient History 2R8 to fiesent _[
= " .a- - - -
g
=
Date Trommest laitiel Power ' Final Pewer Ccaldown? Rs Trip > 20%7 y 4 4/13!B1 Cosidewn from Hot Siby 0 . 0 YES NO S
' c, 4I2541 Peww Mm 100
__ 0 - NO NO = t 82/91 Poww Redsetion -i 100 . 76 NO NO sisi s' 6/21191 Poww Reductan 100 NO . . . _
. 60 NO 3 6f29191 Pewer Reduenos 100 i~ ~ "*"' [
63 NO ND
. S/27191 Power Reduenos 100 72 NO NO - !$,
~
10l9fS1 Shutdown 103 *=- 0 NO N0 ==
- j
* " ~ ~ ' ~~
10123/91 Swtdown 100 0 YES NO E c' 1111/91 Pow w Reduccon 100 80 NO NO 2 :21 11128/01 Pewar Reducuan 100 --- 77 NO NO O b' c 3!SS2 Shutsewn i 100 0 YES
~
L NO _~._~.-
, . - - -_ 5117/92 Power Redmenon 100 73 NO NO ~EE 5/22 S 2 Poww Redue'ian
_ _ . . _ . . _ _ . _ . _ . _ _ 100 . 73 NO NO 22! [m . ._ .. Bl6!92 Power Reduction 100 - 30 NO N0 iEEF ) 7ff~~~~ Bl422 Shutdown ' 100 -- 0 YES NO "i
- N~'~ --
11/191B2 Poww Reducuan 100 80 NO NO M b~- -- 11/21192 Power Reduction 100 30 NO NO C 2f21193 Power Reducnon 100 e 83 NO NO C 3/18193 Powar Reduction 100 80 NO NO , 3/31183 Perw ReductNm 100 9 75 NO ND " 5 - Sf1193 Shutdown ES - - . . 100 0 NO G. g:as==- h*.".,-.....,._
. . , , . . . _ _ _ . . ~ ~ " ' "
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f1AY 06 "33 16:37 GDERATION SUPPORT P.3
""'"A*""
UNIT 2 RCS HEATUPlCOOLDOWN DATE CYCLE DESCRIPTION OF TRANSIENT 8 245-78 I i Precere Het FwmaalTesting g B4&l8 2 Pest Core Hat FenenseelTosong S
~
m-22 78 3 % Rs6ef Veh,e Resejr , 8178 4 Wiped Bennes in 20E2 (Diesel wee inoperabis! Q (2-2-78 5 Replace 'B' Excore Detector ? [2-2478 6 High Clondes in S/8 Dreie & Refil of S!8 b 22378 ' 7 Camdenser Tebe Leak to l(erth Water Box h 2+78 8 2EEA/B Candenser ModWicetan --
&&?9 9 Steam Relief Terting
}w r2-
- 42279 13 S*sem Relief Velve Testine d5 41579 D 8 team Relief Veive Testes A G28-79 Stoes Relief ValveTesting 12 (i
643 78 13 Last power to 2H1&2. 211&2 due to 1stw in switehprar. dg' B479 14 II4CAIB dage ta raadan-r ceased by mdtin!e tube failures Th Anemely Seector interests M-+=a_ bI 10479 15 Dimal M- - :(20R2) 14043 16 I TMI.lassant laarmed Modifme==== 1 4-S80 17 last 500 KVA Trenensemen lines due to westher een ir 'cas 1 443-80 18 Replace CEDM Cet Stack ; 6 5-80 19 Repair Ma's Steen Vent Line Brook 1 RCP Seal repisemeent failevneg pertalics: et effs:ts power ---
)
&7560 20 (ground fault se _500KVline) 8440 21 Cooldews to ciese S.W. Systen Asian cism or: blast --
l 12440 22 Cecidows to fir 2P38 suenos piping kennotatable weld faileral I 3-28 Bl 23 Coeidown for Refusileg Operations Conidewe f or 'A* RCP sesi replacement and 2E4A feedwster
~'
S2741 24 extracuoa line rspair. f Coendown to mode 5 beesne RCS HTU's did net ps:: - a- l 1782 25 remaan time testing. ~ Coeldown te made 5 bar== sf a stsam leek ir. side 4-1642 26 contamount on the "B* SIG blewdewn ime. 42242 27 Cecidews to regner weld leek se 2P32A andde snel sennne ime. 5482 28 Coeidown to replace 2P32C seal 7 2&E2 29 Cooldewn to repair 2F#4A end rep!aes coi stack on CEA 28. Cooldows to Mods 5 becesse sf an unidentifad RCS leak emde mme .
&2042 30 coetemment > 1 spa Begawang cf 2R2.
111642 31 Cooldown for incere reper. -- - - Cooldews te Mode 5 fansplacement cf failed upper gnpper cod __ 121142 32 en CEA-26. 1743 33 Coeldown te Mode 5 due too leelr en 'A' SDC Heat f xdancer. 8-26-83 34 Coeldown to Mode 5 for repeir to 2P32A saaf pressure sensing fee. 02643 35 Conidown to Mode 5 for repairfreole:sment cf battery benk. 1-10-84 38 ' Cooldown to rootsce reactor imod e<xg usakst seals. l 7-20 & } 37 Coeldown to Mede 5 for Stortas Chanas! rectaewnent. ~ - ~ * - B-29 84 { 38 Coeldewn to Mode 5 for RCP ses! rep!seemant. I 142644 l 33 Cooldews to relace coil stack on CEA 7. ----- l _ , = i l Page 1 .
'~
j
t%y 06 '9316:38 GDERATION SUPPORT P.4 UNIT 2 RCS HEATUPlC00LDOWN ""j{"*8 31645 40 CoeWewe fw refueGeg m. taps 2R4. ----:='- 51185 41 Comidews to repair eeelleek as 2P328. - S 9-1345 42 CeeWwe to bde 5 fw *A* sed *C* RCP seed r :' - i.- -
~ " " " ~
f S2845 43 Cee6 dews to hde tw reper to 2P32C seal presswa senase line. [$ 1044b 4 44 Cecidewe to replese 2TE 47114 - ~ ~ " 12R25 45 g Cecidewe to Mode 5 for repaire to artreetwi kne exponeme jemte. g 12/11185 46 Ceeidown to bde 5 for repeir to e CEDM esiL Si14186 47 Cesidewn fer refoeims ostege 2R5. d 111418 6 45 Ceoidows tw repair of 2fW-58. }Le 4I24/87 7l8/87 49 50 Cosidows for repair of oressuruw hester sleeve. Coeidews fw repair of teled weide ne pressurner heatw welis. ' - " h;
~ tu 2I12138 51 Ceeidows tw reteeling outage 2RG.
54l88 52 Coendowe for CEA 8 cosphes 7/24/88 53 Coeidown ta repair leet en 2PSV4533. PZR onde safety vahm. Cooldswe to pwtarm RCP seel pressure seness line MODS d:$ (
- - ~ ~_ , , , , , , ,
8/128 54 sita leak as 'A' RCP. - - 11/m3 l 55 Cecidewe to Mode 5 to receir 2P37C upper esel pinhole leek. 4/18N9 56 Coeidewn to Mode 5 to receir HP Extracuse Ese te 2FIA. N=r
~
&23/E9 57 Coeldews to Mode 5 to repair 2RC.1001 due to high RCS leatage. -
&27/88 58 Cosidows to Mode 5 to esein resair 2RC 1001.
S2G/09 69 Cosidews for refueling estage 2R7. 111/9 0 80 Coeldows ts Mode 5 to teablesheet CFDMCS and FWCS. _ __ 711330 61 Cen&wn to Mode 5 to faciktets repers to presswuw ende safeties. Cooldewn to Mode 5 (from RCS temperature 545 F) when the wrong =- =v . 7/20!90 62 earrie wee discovered to be instened in ene of the code saisbes. Ed 223/91 S3 Coeidewn fw refuerms estase 2R8 """" 4/13S 1 64 CoeWws to Mode 5 forrepair of 2SM4D - - - - - .
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10/23!S1 65 CooWwn fw pressurirm onde safety vahre repleement. #~~~~ 03/09/22 65 Coeldown for Steem Generater Tube Leak 03/0b!92 67 Cooidewn for refue5ng outage 2R9 ~ C ~~ SilS3 G8 Cooldown for 2PS31 S!G tr.soection Ontape I ! p ZCC .~
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timf 06 '93 16:38 GDERATION SLPPORT }
. P.5 i UNIT 2 REACTOR TRIPS miANumiFwam m l OATE DESCRIPTION CYCLE FRACTION :
Sdadded Trip pr TP 1 aM fM Appedir 88. See venamt Reart 1 8 ! r2s40 nm44) 1 ; i.00 . Less of Laud indessen su Condenser Hetwas latsisted MFWP Trip. 4 S , 32840 Rx insped as high pu presswa. (See Trennent Report #2404) 1 1.00 I taas W MFWP commed high pr.pressus Rr trip. j p
- 4243 (see Tramment Roert n4042) 1 ; 1.00 ?
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2 S I L=s of disim pos de m w.ihr imedes As troped .t d Pews = CEDMS. U 6 I (See Trommest Report M4043.) Cysis runst to are sinos it is addressed by less of a h, 4740 eseiset fisw iremment. (Ref: CE apes 00000-PE.140 Fig 4L 0 ' Q.00 gg ; Reesear tripped due to prehlsms with the CEAC red psenes suEconsa en CEA E7.100% j ~ fu l 4-2440 FP.SeeTransmat Report F2615) 1 j 0.00 .O i Reecaer alpped due to problems wnh the CEAC red pennes indissess en CEA 87. (80% ; @M j 42540 FP,see Trenssa Report #24Nel. 1 ; Q.30 61140 Two amadtmasses apansas DNORlLPD trips. (See Trenannt Report F2510.1 1 q 1.00 g[o g ! 1 ! Partini Less si stisite paww due to yesad fault se 500 KV transmanies Isa.191% fr. ~ See Trumasat Rupert F2520L Cysis reest to aere sinos it is addressed by less of E ' S2440 eseisst toe vessiest. (Ref: CE Spee 0000SPE-140 Fig 4L 0 3 (LSD
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Unit gripped due to husen error whss CEAC fl was pleesd in the test neds with CEAC - 15 - 0.03 7740 M in ta lmes made. 93% FP, see Trammast Report F24020.
- 72440 CEA M8 drupped ints the Care. (Ese Trmisuut Repsrt N#22.) 15 iJ10 8440 : Stater anstes waar prehlsm.16es Tramment lisport F24023.) 11 1J10 ,
8440 'A' MpWP res. value taled. ISee Tranment lisport M40-25.) 1 1.00 (
, Unit tripped due to bemma errar whom inserter power was transferred from bettery to i 1 62140 pruary AC seres. (94% FP, Ses Transunt Report F2425 1 ii G.84 Unit tnpped from 27%FP due to loseg *B* RCP R was omsety tnyped. (27% FP, See j iM40 Treamentlisport72 00281 1 - 8.17
= l 104440 Unit tnaped en high SG level caused by suurfeeding. lSee Tranment Report #240-29). 1 ?., 1.00 3
t2&ilo SIB Level trip. (30% FP, See Transest Report #240 30. 12 4.3B , , 122340 Menesi lls tip. (35% FP, See Treament Report M&321. 1 ' 8.35 l i ' 14 4 Dropped CEA82 (See Tranment Report #244D. 1 i 1.00 24 4 R8PT Fatre 18ee Transset Aspert #2442L 1 1.00
. 3-1041 Trip due to low Generater Voltage (See Transest Reports 2443L 1 .; 1.00
- M681 Tfe des to F.W. Cesarel System Prehison (See Tranment Report F2444L 1 1.00 MS8 CPC Chammis I and 2 DNBR Tie.191% FP,8ee Transant Report #24105) 1# ILS1
. l 7741 Roseter trip en steam generator level high. (< 15% FP, See Trennent Report F2446) 1 S 3.15 l Mesusi trip when tosin feedwater flow redeced to a airosit heard failes in tiw *A' WFW t 7G5 Pune speed sename circuit.150% FP. See Transmut Report #2aes) 1 O.50 The reacter tripped en high LPD and low DNBR due to eh penalty factors when PLCEA No.22 dropped.
, 3 748 8 150% FP, See Trennent Report No. 2449) 1 0.50 j Dwial a twhine sysrspeed trip test the turbine tripped en a falso everspeed signal (96% i 84 81 FP.See Tranment Report #2441 1 0.96 , j B88 : PLCEA No. 22 dropped. (2D% FP See Transient Report (2414 1 IL20 e O Page1 ~5 ms; - a _r wom y-. ,
MAY 06 '9316t39 GENERATION ELPPORT P.6 UNIT 2 REACTOR TRIPS "SE""*"" hadvarisetless of main feedwater se bhe oR. The roastor was tripped ,menumer . (95% @ ) 6441 FP,See Traement Bapert #24H2) ?" 11 0.96 8 l EMS leedvertent KSN's elesers. (98% FP, see Trenesent Report #24131 '
' 1 0.98 ,. l Lees of esadoneer vecean des to *B" airnuleths water pens discharge vehre try. [57% i 6 864 ("
- FP.See Transset Resort #2444) 1 ' -
" 0.67 MSN dosere se less of volve operator air supply.
d B-2343 (See Treamset Report F244) 1_ 1.00 o j N twinine tripped se MSR high level trip and the reactor tryped on kw DNBRRugh LPD. d S KH24 (See Treement Report #2417L I~ 1 E 1.00 6 The tsacter tnpped as low DNER and high LPD whea *D RCP tnpped for no apperant i r4 14255 souse. (See Traneant Report 1243L - 1~ 1.00 h2 ' M441 N reacter tnyped en low DNBR due to ==W Th input to the CPCs._ (See Tranment Rymrt F242L) rz 1
= 1.00 jg N unit tnpped as lew 'A'steem generater level due ta en irtadvertent power suppy trip far the leveliestnneantstne hy a technician.
j W , oe 5235 (See Traenset Report (2423J l 1 = 1.00 jy .
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N unit tripped des to a feedwater flew upset, cease undstersesed. , 6 427 & (See Treement Report T2.Bi-24L 1 - i.00
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Meneel reacter tnyped efter both mas feedwater pumps tnpped. Cause unknown. (66% ' - 5-27 8 FP,SeeTreement Report 124-251 : 1 _ 0.86 a N emit tipped en low level' 'A'stente generator due to less of heth MFW pump. (See .i 1204 Traenest Report #242EL E 1 = 140 N emit tripped en low DER during CEAC (2 monthly test. Cause unknown. (See s t2 4 41 Tranomat Report (2427.) i 15 1.00 N uwt tnyped en low DNSR. A turbse geserstar renbad en higli stater comimg T.. temperature ans caused due to isolatios of ACW to both heatsg excheogers. Humes 5 . gag.:. [2214 error.lBl% FP, see Tranneet Report (2428J I. 15 0.91 5 Manuel Rr trip efter both MFW pugs tripped se high discharge prussers. 12442 (83% FP,see Traaneet Report 12421L ! 11 0.83 Manuel Ax trip after both MFW pumps tripped on high dacharge pressurp. 2 12582 (90% FP,See TranmentReport(242 2L 1 2 0.90 Turbine rumbeck to 20 mwe intated by stator cooEng pressuraltamn/Ses limbs. R: tripped en high 81L level 31A8 was initiated en low RCS prsseurs. !' { 3-742 (71%FP,See Tranneet Report #24243) I% 0.71 Artossete Reacter Inp se DNBRlLPD generated by penatty factor from 7CEA's being j
. 4-2842 sesamtchel LD5% FP, See Trsement Reportt2424) r 1 : 0D0 N aut tippal en low Sill level due to boti Main Feedwater Pumps tripping on high ---
[ 5442 dacharge preneura. (84% FP, See Tranannt Report 12426). e
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14 - -tLB4 h reactar tripped des to a low SS level'autated ky the less cf the *B* Hr4ter Drain [ 52142 Pump as Low Heater Dram Tank level (Ses Trans~mnt Report F2424) 1 a 1.00 The reacter tripped on high SIG level due to levd control inrtabil'rties at lew pcwer levels. I 61182 :(9%FP.See Transient P.eport #242-7) N ranctor tripped en lew DNER due to a penalty ft: tor guersted from CCAC-1. AB _ii ~ 0.0E subgroup 20 rods appeared low. Eract cause unknown. 6-1682 (See Traanent Report r242 0) 1 1.00 N rsacter tripped on low QNBR dus to a penaity fa: tar generated frorn PLCEA.28. N _ _ .
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rod dro; ped due to a failure of the CEDM uppw gnpper coils. (EB% FP, see Transient 7-2742 Report (242eL 1- 0.88
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UNIT 2 REACTOR TRIPS
~.z N rsactor tripped dua te a fsilure of the number 12 DC pewer supply for this reactor
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'prstective system CD trip matrix while the number 4 and 8 trip breakara w?are opent.78% S I DJ1-11 442 FP,seeTraanent bewt F24210L 1 -
1~ The tsnetar tnyped as high STG level ibe to level control instahitles at low pewer Irvets. - 11 1642 (16% FP,See Trar.neet Report r24211L 1 0.18- [$ The unit tripped on icw QNER wLib perfwnung a controEed shutdoom. q Q 111842 LD08% FP, Sew Trer.: rest Report (24212J 1 CAB ? __ The wit tripped se a CPC generated low DWBR!hi;h LPD tnp an aH four RPS r6nah 3 m foBerne a spuneus trip of the *B* RCP braaker. 6
- 11-2742 (48% FP,sesTransiest Repxt124213L .
N reacts tripped as low DMER caused from CPC generated pen:lty factors. Irdtis! 1 0.48 h y< cause ws: a dropped red se e reatt of a faSed upper gripper coil - 45 d= 121142 (01.5% FP See Transient Reportf24214L 12 - 1 0E2 . N ust tnpped an low DNER from penalty factora generated in CEAC 1 due to a N gaen 12 22 82 spaneus subgroup rod postica ded.ation.(See Transient Report 124215L 1 1.00 The ucst tripped when a sprinus target rod position bdication in tbs CEACs generated a gg packsd penalty factx which ress:ted in a CPC calculated DNBR et less than 1.24. (See 21443 Tra:mest Repwt #28302.) 1 1.00 Tha tsacts was meemay tnpped waen sa sisctncal fsWt devoleped on S/U E livos 13 and the 22XV windas: of the acts transf anner. (E% FP, see Transiest Raport 12 ! 2115tB3 03) 1 8AE - The reacter tripped from 100% on now DNER, channe!: A and B.The DbR signal vs: S/27S3 pW : a residt af a percerved delta Tc cond.tordSas Transient Report f24344) 1 1.00 The rsector tripped from 100% FP on low DNBR fonowing a twhme trip.;4See Trsasient i 1224l83 Report T2 83 05) __ 1 . 1.00 The reactw tripped from 100% FP on low stesm senerstx level f atewing the loss of + 672 7163 both riais fandwater pumps. (See Transient Report 124346) s 1 1.0') The reactor trip,w! from 68% en low steam generstar level foDowing the loss of both 1 S/24lB3 main feedwater pumps. (See Trrinsient Report 2 83471 - -- 1 'i - --GA8 The rsects tripped from 6% on now steam generstar irrs! d.: ring the transition from EFW $ 9/1l83 to MFW. (See Tranneet Report 243CBI 1 0.0E N reactor trjpped from 4% on low 'A'stram generster pressws. (Sas Tranment
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11334 Report 24442J i' 1 '. CA4 N reacter tripped fross 10% FP no lcw level in "B" strum generstw after the tranction A . 113 1154 frorn EFW to MFW. (See Transiert Report 244- C33 L 1 ; 6.10 N rencor tripped froen 4% FP because of low level in "A" stsa:n genwster. (See _ 3/12iE4 Tranaent Report 2 8445.) 1 6 0A4 The reactor tnpped an tsgh Irvs!in *B'sisam generator dwing a pcwar increase. (57% h FP.See Transient Report 26408.1 1 GE7 05/07/64 The tsactor tripped from 100% on low DNBR caused from CPC genersd penalty f actors 1 [ 1.00 06!17!54 _das to CEA EI drW (Ses Tran:nsnt Report 2S407.1 (he tractar tripped on high levelin 'A'stsam generstw. (10% FP See Transient Report 05/18l64 284DBJ 1 0.10 The reactor was manca!!y tnyped fouowing se ina:trartent switchiig sf vital power . 07/20'84 avertw 2Y11 to an s!tercate source. (See Transient Report 28409) 1 L] 1.00 The reector trioped during heatup on high SG level (0% FP Ses Transiset Repert 2B410.1 1 fLD0 07/2554 i
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t1AY 06 '9316:41 GE1EPATION SUPPORT P.8 l 1 1
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UNIT 2 REACTOR TRIPSE ,, y-.- _-._
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The reactor tnpped on high level ie "B" stearn generster due to manuel foodwater central i
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07128184 error.(See Trenaient Report 28411) 1 1 0.88- 8 l The reacter tryped as low DNBR dus in CIA #4 droppies. (See Traement flaport ! c 1 1.00 0812 8164 26412.) mm __ N reacter tripped from 100% FP en low DNBR when CEA #7 dropped. (See Transeest [ ! 10/2Gl84 Repwt 28413) I 1 1.00 g Tbs esttnpped as a CPC sunliery tnp due te out of tolerance AS:. (7% FP See 9 10/31184 Trenament Report 26414) 1 868 7 3 l
=.- N unit tnaped from 90% FP os low steam generster level due to patial elesere of step 6 11103!84 valves doing stroke testas. (See Tranment Report 284151 F 1 0.90 ! to N erst tripped from 100% FP os high RCS preneurs ieBowng a turbine trip. The __
h~< 02/04/85 turbios trip was due to the less of a sneerster field exateties breaker. 1 1.00 g
- 02/06/55 h umt tnaped from 7% FP on low DNER (-*y ASI trip) durice startup. 1 8.8T ,
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The roector tnpped from 100% FP es new DNER during PPS awtrix testag des to a les 1.00 M
- 07118!S5 et poww to thecrh (See Transeent Report 24411 's 1 a The roector tnpped tres 100% FP en new DNBR. During testing of "D* CPC, a CEA peption deviation eqpud was generated by CEAC 2. ponersting high pensityTactors. (see
[' 1 1 110 07130!B5 Transsent Report 245441 N reacts tripped from 100% FP en low DNBR due to CEAC poseky footws genersted 1.00 08/05.T5 f." , e M sinks in the switchterd. (See Tranaent Report 24Fdl5) 1 N reactet tnpped from 103% FP on high presemuer pressors when a steam generster _ blowdows pump catastrepiecely failed cousing a less of condenser vacuem. (See Troomset Report 245010 1 1.00 08/1325 N reenter tripped from 96% FP on low DNBR fonewng a CEAC F2 card failure. (Ses
'r 1 Q.96 08I16l85 Traenset Report 24M17)
% toector tnpped from 100% FP on high level in T stsom generats wbse T MFW ;
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1 ; 1.00 10lG8;B5 .pumpi- @ .,vekeaa.A N reacter tnpped from 100% FP foDewing loss of condenser vacuurn when the -: 1 i 1.00 10l19185 cir% pose 2P.SB discherpe veno did not class when securing the pump.
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The reacter inpped from 103% FP on low DNBR fol!owing the inadvertect desure of an 02111/88 14Siv. h closure was opparently m==d try a festure of a meta driven reliIy; 1 _a 1.00 - w q a .5 N reactor tripped from 100% FP due to the apparent loss of T RCP. flie cause was # 04/71186 e faulty bght bulb in the pump hands-t$ indication. (See Transient Report 24602) 1 f = 1.00 N reacter tnpped from 100% FP on loss of *D* RCP due to a failed startmg surge ;; 1 f 1.00 04l26/86 capacitor. (See Tressmat Repwt 28503) . The tsactor tripped from 12% FP due to QASI during tha power reduction to begin N 1 -: 8.12 06/13185 ie E , outage 2RS. (See Tramient Report 28804) N reactor tripped from 100% FP due to high levet in the T 14SR. (See Transiset @
' 1.00 1
0G!?4186 Repwt 78605) The reactor tnpped Irors 100% FP due to open in the 'B' phasa of 2X.11 (See Tranerent 1 1.00 02103l87 Report 24741L The reactw tripped from 100% FP due to high PZR pressure f allowing turbine trip on high bearing vibrstion. (See Transient Report 2 87 21. After going critical at 2055.the unit 1 1.00 illit4lB7 4r.be=a due to CPC tnp. ( _ L - - - -
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t1W 06 '9316:42 GOTRATION EUPFORT .e p9 ? e-r l UNIT 2 REACTOR TRIPi~ """""""8 l 0 l The reacter was manusty woped from 100% FP due to a esel seness fm_e f,eRee se 'A' , 8/1/88 RCPlese Treemos: Report 2 8801L F __ . 1 1.00 8 j The reaser tripped from 100% FP des to en inadvertent SLA8 lose Treament Report 2 - 12l1188 0642L
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1~ ---1.00 @ ] The rescaer wipped from 100% FP due a e turtles trip caused by se extracean sesem 5 %
, i 4/18199 Ese repears. 1~ 1.00 4 :
The reester tipped from 100% FP due to high SIG level caused by a FWC8 mailsecues. - 121sil8s " ' , . tsee uTR2 ese2L 1 1.00 4
. 08/28/90 The reester stpped frees 30% FP en law DNBR seused by CEA 29 RSPT Inger to CEAC 1 false. (See UTR.24041L 4 1
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(Out of eneusselThe reester was manuser tripped frosi 20% FP daring power reducten [< ; 4 3Ml90 ts reper surge cepeature se *D* RCP. - 1 : W lC Hoseter tipped from 100% FP se low DER 4ssh LPD siter MSly closed at power des to ; 2: o ' 8/21/90 fagedseleesid. ~.--- 1 " 1.00 . 211/9 1 Rasseur insped from 80% FP due te lose of *B* RCP. 1 - G.as g& m < j j 2!Z2/91 Manand reacter trip from 20% power to basin 2R8 " 1 : 43 gg1 10f9191 Menesd reacter trip trem 20% power for excers . f 1 W $g r - ! 10T23191 Pleased reacter trip from 20% power for reglassment of presswiner code esfory valens 1 - W s. 03/08/82 Pleased reester trip from 20% faI power due to a lealt in the 'A' Steam Generater 1 - W ' 0W05/S2 Flammed reester tnp iram 20% power to bega 2RS = 1 '" - . _W 2 l 5/1/93 Pinesed esector wip frost 20% power to hosto 2PS31 SIS inspecuan Detage 1 .. W ! TOTALS c__ 153 73J2 1 _ . . I t. 4_. menset--. , _. ._-.u , . _ _ . C. Wis m en = - e _ _ _ _ _ _ _ 3 c , u.-. -- p- . _ _ _ _ : g g ,_ > c... _
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, , PFrf 06 '9316:43 GEtERATION SLPPORT h P.13 F
UNIT 2 TURBINE TRIP WITH DELAYED ;-- [ REACT 0lfTRIP"85"
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DATE CFCiE DESCRIPTIONC- asmen; 9/28190 /1 Turbias trip es isw eendenser veamum at 82% FP fouswed by a mammel reestar trip. ---- a !
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