ML20039F866

Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for Waterford Reactor Vessel.

ML20039F866
Person / Time
Site:	Palisades, Waterford
Issue date:	12/31/1981
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML13308A045	List: ML13308A044 ML14092A301 ML17297B168 ML20039F858 ML20039F860 ML20039F861 ML20039F862 ML20039F863 ML20039F866 ML20039F867 ML20039F868
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-I, NUDOCS 8201130489
Download: ML20039F866 (7)

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CEN 189 APPENDIX 1 7

EVALUATION OF PRESSURZED THERMAL SH0CK EFFECTS DUETO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE WATERFORD REACTOR VESSEL Prepared for THE LOUISIANA POWER AND LIGHT COMPANY NUCLEAR OWER SYSTEMS DIVISION PI ' POWER Bliiil SYSTEMS g c7 coueusrionenoineenino.ine

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1 LEGAL NOTICE

. THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COM8USTION ENGINEERING i 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.

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-ABSTRACT This Appendix to CEN-189 provides the plant-specific evaluation of pressurized thennal shock effccts due to small break LOCA's with extended loss of feedwater for the Waterford 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 fM1 plant life.

I

CEN-189 Appendix 1 TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT PURPOSE Il II.

I2. SCOPE Il I3. INTRODUCTION Il I4, THERMAL HYDRAULIC ANALYSES I2

15. FLUENCE DISTRIBUTIONS 12

16. MATERIAL PROPERTIES 13

17. VESSEL INTEGRITY EVALUATIONS 15

18. CONCLUSIONS 19 ii

11.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized themal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the Waterford reactor vessel.

12.0 -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 Waterford reactor vessel.

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Other C-E NSSS reactor vessels are reported in separate Appendices.

13.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Louisiana Power and Light for their use in responding to Item II.K. 2.13 of NUREG-0737 for the Waterfard 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 15 of this Appendix reports the plant-specific fluence distributions _ developed as described in Chapter 5.0 of the main report. Chapter I6 reports the plant-specific material properties and change of properties -due to irradiation, based on the methods of Chapter 6.0 of the report. Chapter 17 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to a set of material properties which are conservative with respect to the plant sepcific properties reported in Chapter I6.

This additional conservatism was not remosed because of the favorable results.

Il

I4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to perform 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 w icn could be eliminated by performing more detailed plant-specific thermal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.

I.5.0 FLUENCE DISTRIBUTION The Waterford Station is not yet in operation and 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 pre-diction was used to estimate end of life material properties.

Also, in order to evaluate the sensitivity of the fluence predic-tion value, material properties were also determined assuming an end of life fluence twice the FSAR prediction value.

APPENDIX I h"ATERFORD UNIT #3 I.6 MATERIAL PROPERTIES The chemistry and initial (pre-irradiation) toughness properties of the Waterford Unit #3 reactor vessel beltline materials are summarized in Table I6-1. The most controlling material in terms of residual chemistry (copper and phospho rus) and initial RT is plate M-1004-2 from the lower shell course.

NDT The predicted RTNDT shift based on the maximum, design fluence, 3.68 x 1019n/cm2 (E>1MeV) at the inside surface of the reactor vessel, is 77F using Regulatory Guide 1.99. This will result in an adjusted RTNDT at end-of-life (32 effective full power years) of 99F at the vessel inside surface. If the design fluence was 2

increased by a factor of two to 7.36 x 10 n/cm , the RTNDT shift is predicted to be 109F for an adjusted RTNDT of 131F.

G I6-1 13

4 TA3LE 16-1 WATERFORO UilIT #3 REACTOR VESSEL MATERIALS d

Product . Material Drop Weight Initial Chemical-Content'(%)

Form Identification NDTT ( F) RTNDT ( F) Nickel _ Copper Phosphorus Plate M-1003-1 -30 .30 0.71 0.02 0.004 Plate M-1003-2 -50 -50 0.67 0.02 0.006 Plate M-1003-3 -50 -50 0.70 0.02 0.007

-50 -16 10.62 0.03 0.006

. Plate M-1004-1

-20 22 0.53 0.03 0.005 Plate M-1004-2

-30 16 0.62 0.03- 0.007 Plate M-1004-3 Weld 101-124 A,B,&C^ -60 .

-60 0.96 0.02 0.010 b -80 <0.20 0.03 0.007 Weld 101-142 A,B,&C -80 c -70 to -80 0.16 0.05 0.003 Weld 101-171 -70 to -30 C.

a Intermediate shell course longitudinal seam weld b Lower.shell course longitudinal seam weld' c Intennediate-lower shell girth weld d Plate RTsnT determined using Branch Technical Position MTEB 5-2; weld RT NDT determined in accorda'nce with ASME Code, Section III~, NB-2300

f.750 Waterford 3 Vessel Integrity The fracture mechanics analysis is performed uing upoer bound data

-for fluence and material properties in the Waterford .3 vessel. The peak _ vessel fluence is considered to occur at the point of maximum RT The material toughness properties K and NDT. IC K

h are determined from the calculated temperatures for the SBLOCA

+ LOFW transients using the method described in Section 7.3.3 and predicted RT NDT values through the depth of the wall. Critical crack depth diagrams are constructed from the applied Ky vs crack depth curves at the mid-core level of tne vessel and the calculated material toughness curves. In this manner the integrity of the Waterford 3 vessel is evaluated for the SBLOCA + LOPA transients.

X.7.1 Sumary of Physics and Material Data Input to Fracture Mechanics Analysis A nominal design fluence value of 3.68 x 10 19 n/cm2 (E >l MeV) was used to approximate the end-of-life fluence for the Waterford 3 vessel, as well as a conservative upper bound of 7.26 x 10 19 n/cm 2 or double the predicted end-of-life value. The peak fluence is considered to be uniform around the vessel. A conservative radial fluence attenuation was used such that: -

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[ = exp (-8.625 in. x .33 in. I) -

(a/w)

F g

= exp (-2.85) -

(a/w) where F = point fluence in wall F,

= peak fluence at surface a/w = fractional wall depth Upper bound materials data were used to conservatively envelope all plate and weld materials, which are as follows:

PCT. Cu = .10 PCT. P = .010 Initial RT riDT

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The shift in the value of the RT NDT was determined using the method of Reg. Guide 1.99. This produces an end-of-life prediction for the surface RT NDT of 174 F using the nominal design fluence. A predicted surface RT NDT value cf 230 F is determined for a fluence double the nominal design fluence.

I.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 the report. The fracture mechanics analyses were performed using upper bound properties for the Waterford 3 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 peak fluence and material properties.

Figure I.7-1 shows the critical crack-depth diagram for a nominal design fluence of 3.68 x 10 19 n/crt 2 The calculated shifts in RTNDT are relatively low, and for this transient 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 I.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 greatly exceeding the nominal design fluence.

I.7.3 Conclusion These results demonstrate that the integrity of the Waterford 3 vessel would be maintained throughout the assumed life of the plant for the SBLOCA + LOFW transient with recovery of feedwater.

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18.0 CONCLUSION

S This Appendix to CEN-189 provides the results of analytical ' evaluations of pressurized thermal shock effects on the Waterford reactor vessel for cases of a SBLOCA + LOR 4, in response to the requirements of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based an remedial actions to prevent inadequate core cooling:

1. SBLOCA + LOFW + PORV's opened after 10 minutes

2. SBLOCA + LOR 4 + Aux. FW reinstated after 30 minutes Thermal-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 using linear elastic fracture mechanics methods to determine 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 is 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 specific 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 + LOPW transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.

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