ML20039F860

Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for San Onofre 2 & 3 Reactor Vessels.

ML20039F860
Person / Time
Site:	Palisades, San Onofre
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-H, TAC-59981, TAC-59982, NUDOCS 8201130483
Download: ML20039F860 (14)

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Sv' EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECT DUE TO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER

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FOR THE SAN ON0FRE 2 & 3 REA' TOR VESSELS Prepared for THE SOUTHERN CALIFORNIA EDISON COMPANY NUCLEAR VER SYSTEMS DIVISION 8e01130 yS3 Pi ' POWER Emmid SYSTEMS CCMBUSTION ENGINEERING INC

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LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COM8USTION 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 DISCLOSEO IN THIS REPORT.

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ABSTRACT This Appendix to CEN-189 provides the plant-specific evaluation of pressurized thennal shock effects due to small break LOCA's with extended loss of feedwater for the San Onofre 2 & 3 reactor vessels. 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.

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1 CEN-189 Appendix H TABLE OF CONTENTS SECTION . TITLE PAGE ABSTRACT H1 Hl. PURPOSE H1 H2. SCOPE H1 H3. INTRODUCTION H4, THERMAL HYDRAULIC ANALYSES H2 FLUENCE DISTRIBUTIONS H2 H5.

H6. MATERIAL PROPERTIES H3 VESSEL INTEGRITY EVALUATIONS H6 H7.

H8, CONCLUSIONS H10

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Hl.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; San Onofre 2 & 3 vessels.

H2.0 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 San Onofre 2 & 3 reactor vessels.

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

H

3.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Southern California Edison for their use in responding to Item II.K.2.13 of NUREG-0737 for the San Onofre 2 & 3 reactor vessels.

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 H5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter H6 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 H7 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 specific properties reported in Chapter H6.

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

H4.0 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 which 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 conclu: ion achieved. .

H.5.0 FLUENCE DISTRIBUTION The San Onofre 2 & 3 Units are not yet in operation and have 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 evaluate the sensitivity of the fluence pre-diction value, material properties were also determined assuming an end of life fluence twice the FSAR prediction value.

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'RPPENDIX H SAN ONOFRE UNITS #2 and #3 H.6 MATERIAL PROPERTIES The chemistry and initial (pre-irradiation) toughness properties of the San Onofre Units #2 and # 3 reactor vessel beltline materials are summarized in Tables H6-1 and H6-2, respectively. 'The most controlling material in terms of residual chemistry (copper and phosphorus) and initial RTgyp is plate C-6404-2 from the

' intermediate shell course for Unit #2 and plate C-6802-1 from the intermediate shell course for Unit #3. The predicted RTNDT. shift based on the maximum design fluence, 3.68 x 1019n/cm2 (E>lMev) at the inside surface of the reactor vessel, is 134F for Unit #2 and 77F for Unit #3 using Regulatory Guide 1.99. This will result in an adjusted RTNDT at end-of-life (32 effective full power years) ~ of 150F and ll7F at the vessel inside surface for Units #2 and #3, respectively.

If the design fluence was increased by a factor of two to 7.36-x 10 19n/cm ,.the 2

RTUDT shift-is predicted to be 190F and 109F, resulting in adjusted RTNDT values of 206F and 149F for. Units #2 and #3, respectively.

.H6-1 H3.

TABLE 116-1 SAN ON0FRE UNIT #2 REACTOR, VESSEL MATERIALS.

Product Material Drop Weight Initial Chemical ~ Content (%)

Identification NDTT ( F) RTNDT ( F) Nickel _ Copper Phosphorus Fo rm -

Plate C-6404-1 -30 16 0.55 0.10 0.009 Plate C-6404-2 -20 16 0.53 0.10 0.010.

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., Plate C-6404-3 -20 13 0.53 0.10 0.009 Plate C-6404-4 -10 14 0.60 0.09 0.006 Pla te C-6404-5 -20 2 0.62 0.11 0.006 Plate C-6404-6 -10 -10 0.57 0.10. 0.007 Weld 2-203 A,B,&Ca -60 -60 0.95 .0.03 0.010 We ld 3-203 A,B,5Cb -50 -50. 0.12 0.06 0.011 c -60 -50' O.29 0.07 0.009 Weld 9-203 a Intermediate shell course longitudinal seam welds b Lower shell course longitudinal seam welds c Intermediate to lower shell . girth weld l

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i TABLE 116-2:

SAN ON0FRE UillT #3 i REACTOR VESSEL MATERIALS ,-

Product Material Drop Weight Initial Chemical Content (%)

Form Identification NOTT (*F) RTNDr ( F) Nickel Copper Phosphorus Plate C-6802-1 -20 +40 0.53 0.06 0.002 Plate C-6802-2 -20 0 0.59 0.05 0.007 g Plate C-6802-3 -10 +20 0.59 0.06 0.002 Plate C-6802-4 -30 +5 0.56 0.05 -0.007

-Plate C-6802-5 0 +10 0.53 0.04 0.010 Pla te C-6802-6 -40 +20 0.61 0.06 0.006 8 0.05 0.011 Held 2-203 A,B,&C -40 -40 0.21 b 0.04 0.012 tield 3-203 A,B,5C -70 -10 0.21 tield 9-203c -50 -50 0.05 0.06 0.010 E

a Intermediate shell course longitudinal seam welds b Lower shell course longitudinal seam welds c Intermediate-to lower shell girth weld O

H.7.0 SONGS 2 & 3 Vessel Integrity The fracture mechanics analysis is performed upper bound data for fluence and material properties in the SONGS 2 & 3 vessels. The peak vessel fluence is considered to occur at the point of maximum RT The material toughness properties K IC and NDT.

Kg , 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 the vessel and the calculated material toughness curves. In this manner the integrity of the SONGS 2 & 3 vessels are evaluated for the SBLOCA + LOFW transients.

l H.7.1 Summary of Physics and Material Data Input to Fracture Mechanics Analysis A nominal design fluence value of 3.68 x 10 19 n/cm 2 (E >l MeV) was used to approximate the end-of-life fluence for the SONGS 2 & 3 vessels, as well as a conservative upper bound of 7.36 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:

[ = exp (-8.625 in. x .33 in. 1) ,

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

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= peak fluence at surface

= fractional wall depth a/w 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 NDT = 0%

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The shift in the value of the RTNDT was determined using the method of Reg. Guide 1.99. This produces an end-of-life prediction for the surface RT of 174 F using the nominal design fluence. A predicted ,

t1DT surface RT value of 230 F is determined for a fluence double the f1DT nominal design fluence.

H.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 SONGS 2 & 3 vessels 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 H.7-1 shows the critical crack depth diagram for a nominal design fluence of 3.68 x 10 19 n/cm2 . The calculated shifts in RT riDT 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 H.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.

H.7.3 Conclusion These results demonstrate that the integrity of the SONGS 2 & 3 vessels would be maintained throughout the assumed life of the plant for.

SBLOCA + LOFW transient with recovery of feedwater.

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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 SONGS 2 + 3 reactor vessels for cases of a SBLOCA.+ LOFW, in response to the requirements.of Item 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 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 detennine 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 croner-ties to define vessel capability versus plant-life. The results of i

the generic LEFM analyses were evaluated using the plant specific j 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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