View source for IA-86-569, Safety Evaluation Concluding That Ferritic Components in Containment Pressure Boundary Meet ASME Code Section III Fracture Toughness Requirements & That GDC 51 Requirements Satisfied.Salp Rept Encl

{{Adams
| number = ML20206L969
| issue date = 09/16/1985
| title = Safety Evaluation Concluding That Ferritic Components in Containment Pressure Boundary Meet ASME Code Section III Fracture Toughness Requirements & That GDC 51 Requirements Satisfied.Salp Rept Encl
| author name = 
| author affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| addressee name = 
| addressee affiliation = 
| docket = 05000423
| license number = 
| contact person = 
| case reference number = FOIA-86-569
| document report number = NUDOCS 8608200420
| package number = ML20204G513
| document type = SAFETY EVALUATION REPORT--LICENSING &amp; RELATED ISSUES, TEXT-SAFETY REPORT
| page count = 4
}}

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ATTACHMENT 1 NORTHEAST UTILITIES MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423' MATERIALS APPLICATION SECTION MATERIALS ENGINEERING BRANCH 6.2.7 Fracture Prevention of Containment Pressure Boundary In a previous SER input we indicated that ferritic materials that are used in the containment pressure boundary will be reviewed to the fracture tough-ness criteria for Class 2 components identified in the Summer 1977 Addenda of Section III of the ASME Code.        For Class 2 components, the fracture toughness criteria in the Summer 1977 Addenda of Section III of the ASME Code permits the materials to be either Charpy V-notch tested at or below the Lowest Service Temperature, evaluated to the nil-ductility transition 4
temperature requirements of Table NC-2311(a)-1 of the ASME Code, or evaluated using the fracture mechanics methods contained in Appendix G of the ASME Code.
Ferritic materials that are in the Millstone-3 containment pressure-boundary were procured to earlier fracture toughness criteria than those in the Summer 1977 Addenda of the ASME Code.              Hence, many materials were not Charpy V-notch tested at or below'the Lowest Service Temperature.              To demonstrate that these materials meet the review criteria, the applicant used the fracture toughness data presented in NUREG-0577, " Potential for p 00620                                                                                                      l HShpf $'0 esog,g 6 369 PDR

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                                                            -g-Low Fracture Toughness and lamellar Tearing on ,PWR Steam, Generator and Reactor Coolant Pump Supports," USNRC, October 1979 and ASME Code Section III, Summer 1977 Addenda, Subsection NC.         This data indicates that all materials meet the nil-ductility transition temperature criteria of Table NC-2311(a)-1 except for ferritic materials in the feedwater line.
The ferritic materials in the feedwater line were evaluated using the fracture mechanics methods in Appendix G of the ASME Code. The licensee used a lower bound reference stress intensity factor (26.78 ksi /Tii.) for determining the allowable material fracture toughness.         According to Appendix G, the reference stress intensity value used in the analysis would be applicable for ferritic material at 180 F below the materials nil-ductility I
transition temperature. Additional fracture toughness data for materials with similar composition and heat treatment as the Millstone 3 feedwater materials is reported in a text by Rolfe and Barsom titled, " Fracture and Fatigue Control in Structures, Applications of Fracture Mechanics" (Prentice-Hall,1977).      This data indicates that the reference stress intensity value assumed in the Appendix G fracture mechanics analysis is conservative. The crack sizes assumed in the evaluation were greater than that permitted during the preservice examination of the component and allowed for flaw growth in service.        The fracture mechanics analysis

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indicates that the ferritic materials in the feedwater li,ne would meet the safety margins recommended in Appendix G of the ASME Code. Additional fracture mechanics analysis performed by the licensee indicates that the critical crack size for brittle fracture would be greater than twice the depth used in the Appendix G analysis.
Based on our review of the available fracture data and material fabrication histories, the use of correlations between metallurgical characteristics and material fracture toughness, and fracture mechanics analysis performed by the licensee, we conclude that the ferritic components in the Millstone 3 containment pressure boundary meet the fracture toughness requirements that are specified for Class 2 components by the 1977 Addenda of Section III of the ASME Code. Compliance with these Code requirements provides reasonable
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assurance that the Millstone 3 reactor containment pressure boundary will behave in a nonbrittle manner, that the probability of rapidly propagating fracture will be minimized and that the requirements of GDC 51 are satisfied.
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SALP REPORT                                                   i 1
PLANT: Millstone Nucl'ar                                                                                                                       e Power                   LICENSEE:   Northeast Utilities Station, Unit 3                                                                                                                   REVIEWER:   B. Elliot DOCKET N0.:                                                                               50-423                                                                         LICENSING ACTIVITY: General Design Criteria 51 EVALUATION CRITERIA                                                                                                                                         RATING           REMARKS
: 1. Management Involvement                                                                                                                                    1 and Control in Assuring Quality
: 2. Approach to Resolution                                                                                                                                     1 of Technical Issues from a Safety Standpoint
: 3. Responsiveness to NRC                                                                                                                                     2 Initiatives
: 4. Enforcement History                                                                                                                                      NA
: 5. Reporting and Analysis                                                                                                                                   NA of Reportable Events
: 6. Staffing                                                                                                                                                  1 Overall                                                                                                                                                       1
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