ML20039F868

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Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for AR Nuclear One- Unit 2 Reactor Vessel
ML20039F868
Person / Time
Site: Palisades, Arkansas Nuclear  Entergy icon.png
Issue date: 12/31/1981
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML13308A045 List:
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-G, NUDOCS 8201130492
Download: ML20039F868 (14)
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{{#Wiki_filter:.- p su2' ???es?uo EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECTS DUETO SMALL BREAK LOCA'S WITH LDSS OF FEEDWATER FOR THE ARKANAS NUCLEAR ONE--UNIT 2 REACTOR YESSEL Prepared for ARKANSAS POWER AND LIGHT COMPANY NUCLEAR /ER SYSTEMS DIVISION Pi POWER Bliiiil l SYSTEMS i 8 2 o 1 i <> n yi F C M USTI N ENGINEERING. INC

j i I LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT. e ev*w - ~ w--e-s- ~ ~- e o ---s,--e-- ,-w e ,m- - - +, -,,,y- -,-,-.---e w----r-,.

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ABSTRACT ~ This Appendix to CEN-189 provides the plant-specific evaluation of pressurized thermal shock effects due to small break LOCA's with extended loss of feedwater for the ANO-2 reactor vessel. It is concluded that crack initiation would not occur for the transients considered for more than 32 effective full power years, which is assumed to represent full plant life. i

I CEN-189 Appendix G TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT Gl. PURPOSE -Gl' G2. SCOPE-131 1 G3. INTRODUCTION Gl G4. THERMAL HYDRAULIC ANALYSES' Gl G5. FLUENCE DISTRIBUTIONS G2 'G6. MATERIAL PROPERTIES G3 G7. VESSEL INTEGRITY EVALUATIONS G5 G8 CONCLUSIONS -G9 I l t i if

Gl.0. PURPOSE This Appendix provides the plant-specific evaluation of pressurized thermal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the ANO-2 reactor vessel. G2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA + LOFW transients presented in CEN-189, as applied to the ANO-2 reactor vessel. Other C-E NSSS reactor vessels are reported in separate Appendices. G

3.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Arkansas Power and Light for their use in responding to Item II.K.2.13' of NUREG-0737 for the ANO-2 reactor vessel. This Appendix is intended to be a companion to the CEN-189 report. The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter G5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter G6 reports the plant-specific material properties and change of properties due to irradiation, l based on the methods of Chapter 6.0 of the report. Chapter G7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter G6. G4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to perfom the plant-specific vessel evaluation reported in this Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used j which could be eliminated by performing more detailed plant-specific themal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved. G1

.G.5.0 ; FLUENCE DISTRIBUTION Arkansas Nuclear One - Unit 2 is in operation, but has not yet completed a surveillance capsule evaluation. Since the vessel beltline materials are' low copper content,. detailed fluence profiles were not -necessary for demonstration of acceptable PTS capability. Accordingly, the FSAR end of life peak fluence prediction was used to ' estimate end 'of life material properties. Also, in order' to ev'aluate the sensitivity of the fluence predic-tion value, material properties were also detennined assum-ing an end of life fluence twice the FSAR prediction value. G2

4 t APPENDIX G ARKANSAS NUCLEAR ONE UNIT #2 .G.6 MATERIAL PROPERTIES The chemistry and initial (pre-irradiation) toughness properties of the Arkansas Nuclear One - Unit #2 reactor vessel beltline materials are summarized in Table G6-1. The most controlling material in terms of residual chemistry (copper and phosphorus) and initial RT based on Regulatory Guide 1.99* is NDT plate C-8009-1 from the intermediate shell course. The predicted RTNDT shift based on the maximum design fluence, 3.47 x 1019n/cm2'(E>lMeV) at the~inside -surface of the reactor vessel'is 168F using Regulatory Guide 1.99. This will result'in an adjusted RTNDT at end-of-life (32 effective full power years) of 173F at the vessel inside surface. If the' design fluence was increased by a - 19 2 factor of two to 6.94 x 10 n/cm, the RTNDT shift is predicted to be 237F;for an adjusted RTNDT of 242F. i i N i L l t-

  • Note that plate C-8009-3 is the plate selected for the surveillance program i

on the basis of the C-E design curve presented in the ANO-2 FSAR. i l i l I l G6-1 i G3-

TABLE G6-1 ARKAftSAS i!UCLEAR ONE - UNIT #2 REACTOR VESSEL f1ATERIALS Product Material Drop Weight 'Initiale Chemical Content-(%) Form Identification NDTT (*F) RTNDT ( F) Nickel Copper Phosphorus Pla te C-8009-1 -30 5 0.63 0.12 0.010 Plate C-8009-2 0 10-0.59 0.03 0.009 Plate C-3009-3 -10 35 0.60 0.03 0.009 c Pla te C-8010-1 -20 10 0.59 0.03 0.006 Plate C-8010-2 -30 -20 0.66 0.07 0.003 Plate C-8010-3 -30 -20 0.65 0.07 0.003 d Weld 2-203 A,B,aC" ' H/A <10 <0.20 0.05 0.013 Ueld 3-203 A,B,AC il/A < 10 - <0.20 0.04 0.011 b c f Neld 9-203 -40 -40 0.03 0.04 0.004 { ? Intermediate shell course. longitudinal seam weld a b Lower shell course lonqitudinal seam treld Intermediate to lower shell girth weld c d fl/A - flot Available f Detennined usinq Branch Technical Position MTEB 5-2 e l f Surveillance Program Data l 1 i

G.7.0 AN0-2 Vessel Integrity The fracture mechanics analysis is performed using upper bound data for fluence and material properties in the ANO-2 vessel. The peak vessel fluence is considered to occur at the point of highest RT The material toughness properties K and K are detennined NDT. IC la from the calculated temperatures for the SBLOCA + LOFW transients using the method described in Section 7.3.3 and the predicted j RT values through tne depth of the wall. Critical crack depth D NDT diagrams are constructed from the applied K vs crack depth curves g at the mid-core level of the vessel and tha calculated material toughness curves. In this manner the integrity of the ANO-2 vessel is evaluated for the SBLOCA + LOFW transient. J..' G.7.1 Sumary of Physics and Matarial Data Input to Fracture Mechanics kna19s i; i 19 2 d, (( A nominal design fluence value of 3.47.x 10 ri/cm (E >1 5eV) h s used.~ to approximate the end-of-life fluence for the ANO-2 vessel, as' well-as a conservative upper bound of 6.94 x 1019 h m or double the predic,tyt' 2 ~' The peak fluence is coiisi'dered to be uniNrin arohnd <[f{ end-of-life value. the vessel. A conservative radial fluence < attenuation was used sucfi that: ~ -s. ~ F ,// + l (-8.625 in x.33 'in.f ) (a/w) 'j ,/ exp = F s. /. o ~ q r, (a/w) ~ exp (-2.85) = ~ rf point fluence in wall where F = peak fluence at surface F = g a/w fractional wall depth = / ,e / Controlling plate material properties were used in the analy' sis, ~~' which are as follows: /~ ; .12 PCT. Cu = .s / PCT. P _= .010 'i Initial RT NDT-The shift in the value of the RTNDT was determined using the method #,e

ar e .q,, r ' ~ ' (. of Reg. Guide 1.99. This produces an end-of-life prediction for the surface RT f 173 F using the nominal design fluence. A predicted NDT surface RT value of 242 F is determined for a fluence double the NDT nominal design fluence. G.7.2

Results of Fracture Mechanics Analysis for SBLOCA + LOFW Restoration of Feedwater (Case-5)

The stress analysis for this case is presented in Section 7.8.2 of V the report. The fracture mechanics analyses were performed using y upper bound properties for the ANO-2 vessel and conservative end-of-life - fluence levels. The critical crack depth diagram is constructed using ~* the stresses in the transient at the mid-core level coincident with the h '7 .v peak fluence and material properties. Figure G.7-1 shows the critical-3.47x 10 ' n/cm, I 2 /'E crack depth diagram for a nominal design fluence of The calculated shifts in RT are relatively low, and for this transient NDT loading condition the initiation toughness level is not exceeded. Therefore, no crack initiation would occur for this combination of loading, fluence, and material properties. Figure G.7-2 shows the critical crack depth diagram for the same transient loading and upper bound material properties, but twice the nominal design fluence. From the figure it is apparent that no crack initiation would occur for this transient even with fluence levels 4 greatly exceeding the nominal design fluence. G.7.3 Conclusion These results demonstrate that the integrity of the ANO-2 vessel would be. maintained throughout the assumed life of the plant for SBLOCA + LOFW (, ~ transient with recovery of feedwater. s p 1. / 4 [$ , h "* G6

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8.0 CONCLUSION

S This Appendix to CEN-189 provides the results of analytical evaluations of pressurized thermal shock effects on the ANO-2 reactor vessel for cases of a SBLOCA + LOFW, in response to the requirements of Iteni II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based on remedial actions to prevent inadequate core cooling: 1. SBLOCA + LOFW + PORV's opened after 10 minutes 2. SBLOCA + LOFW + Aux. FW reinstated after 30 minutes Themal-hydraulic system transient calculations were performed on a reference-plant basis, as reported in CEN-189 with the parameter variations over the range representing all operating plants. Four different cases were analyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of the two different scenarios was analyzed usino linear elastic fracture mechanics methods to detemine the critical crack tip stress intensity values for comparison to plant specific materials properties at various times in plant life. The effect of the warm prestress phenomenon 1. identified where applicable for each transient, and credited where appropriate. In this Appendix, the results of plant specific peak neutron fluence + predictions are superimposed on plant sepcific material proper-ties to define vessel capability versus plant life. The results of the generic LEFM analyses were evaluated using the plant specific material properties. It is concluded that crack initiation would not occur due to the SBLOCA + LOFW transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.

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