ML20216E740
| ML20216E740 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/23/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20216E656 | List: |
| References | |
| TAC-60075, NUDOCS 8706300744 | |
| Download: ML20216E740 (5) | |
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- p SAFETY EVALUATION BY THE OFFICE OF NUCLFAR REACTOR REGULATION 1
RELATFD TO AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. NPF-2 ALABAMA PCH N COMPANY j
JOSEPH M. FARLN NUCLEAR PLANT, UNIT NO.1 DOCKET NO. 50-348 INTRODUCTION In c letter from R. P. Mcdonald to S. A. Varga dated October 25, 1985, the Alabama Power Company ((the licenseel requested changes to the, Joseph M. Farley Nuclear Plant, linit 1 Farley-11 Heatup/Cooldown Curves and s.spporting bases.
The staff reviewed the licensee's submittal and det' ermined that the effect of neutron irradiation on all be'tline mat'e'rlais had not been arrassed by the licensee. The staff's concerns are documented in a letter from E. A. Reeves to R. P. Mcdonald dated June 16, 1986.
In response to this letter, the licensee 3
provided a revised set of heatup and cooldown curves, which were to be appli-l cable for 16 effective full power years (EFPY). The revised curves and their btses were submitted in a letter from R. P. Mcdonald to L. S. Rubenstein dated f'
September 29, 1986. The curves and bases are to be contained in Figures 3.4-2 and 3.4-3 and Bases Section 3/4.4.10 of the Farley-1 Technical Specifications.
DISCtlSSION Heatup/Cooldown curves must be calculated in accordance with the requirements of Appendix G of 10 CFR 50, which became effective on July 26, 1983.
Appendix G of 10 CFR 50 requires that the reactor vessel beltline and closure flange region materials meet the safety margins of Appendix G of the ASME Code Section III. To calculate pressure-temperature limits in accordance withi these requirements, the effect of neutron irradiation, boltup, pressure and thermal stresses on the limiting reactor vessel beltline and closure flange ree,fon materials must be estimated. The effect of neutron irradiation on the Farley-1 beltline materials is documented in' Westinghouse Report WCAP-10934, Rev. 2 dated June 1986, which is in Attachment 2 to the licensee's submittal dated September 29, 1986. The effect of boltup, pressure and thermal stresses on the reactor vessel closure flange region are documented in letters from R. P. Mcdonald to S. A. Varga dated June 18, 1984, and October 25, 1985.
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EVAltlATION The methods recommended by the NPC staff for calculating the effect of neutron irradiation damage are documented in Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."
The relationships documented in Regulatory Guide 1.99 were empirically derived from materials that were irradiated in commercial nuclear reactor surveillance capsules. The most pertinent empirical relationships are contained in Revision 2 to the guide. Revision 2 has been reviewed by the NPC staff and published for comment. The licensee has used the methods recommended in Regulatory Guide 1.09, Revision 2, to calculate the effect of neutron irradiation on the Farley-1 beltline materials.
Neutron irradiation damage is measured by an increase in a material's reference temperature. 1he value of the reference temperature that results from neutron irradiation damage is called the material's adjusted reference temperature, ART. The limiting ART was used to calculate pressure-temperature limits for the Farley-1 beltline materials. These limits were calculated in accordance with the reovirements in Appendix G of the ASME Code Section III. The NRC staff has evaluated these llmits using the calculation methods recommended in Standard Review Flan (SRP) 5.3.2, " Pressure-Temperature Limits."
The stresses in the closure flange region resulting from pressure, thermal effects, and boltup were calculated by the licensee using finite element analysis. The closure head and vessel flange geometry used in the finite element analysis was modelled for a typical 4-loop reactor vessel. The Farley-1 plant is a 3-loop reactor vessel. The geometry of the closure flange region in the Farley-1 reactor vessel is slightly different than that of the typical 4-loop reactor vessel. To account for these differences, the licensee used the computation method of Reference 1 to perform a stress analysis of the Farley-1 vessel based on the finite element analysis and an analytical comparison of critical dimensions of the two types of vessels. Their analysis indicates that the typical 4-loop reactor vessel and the Farley-1 reactor vessel have essentially equivalent pressure and boltup stresses at the critical closure flange region. Hence, the stresses from boltup and pressure used for i
the typical 4-loop plant were used in the fracture mechanics evaluation for Farley-1. The stresses at the critical closure flange region resulting from thermal conditions during heatup or cooldown of the Farley-1 vessel were l
detemined by the computation method in Reference 1 to be significantly less than i
those calculated for the typical 4-loop plant.
Fracture mechanics evaluations at three discontinuity locations in the closure i
flange region were performed in accordance with the methodology in Appendix A of ASPE Code Section XI.
In this analysis the licensee used all the safety factors reouired by Appendix G of the ASME Code, except for the Code recommended flew size, to determine the closure flange location that would be
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considered the critical location. The location with the highest stress intensity factor after applying safety margins was considered the critical closure flange location. The critical location was detemined to be the outside surface at the discontinuity between the flange and upper shell of the reactor vessel.
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The postulated flaw size recommended by Appendix G of the ASME Code was used l
for evaluating the beltline region, but was not used in evaluating the closure flange region. The postulated flaw size recommended by Appendix G has a depth of i the section thickness (i T) and a length of 11 times the section thickness. The section thickness at the critical flange location for Farley-1 is 9.125 inches. Appendix G of the ASME Code indicates that smaller defect sizes may be used on an individual case basis, if a smaller size of maximum postulated defe,:t can be assured. The postulated defect used in the licensee's analysis was a 0.625 inch deep by 3.75 inches lono surface flaw. The licensee's justification for using a smaller flaw size in evaluating the closure flange region than that used in evaluating the beltline region is that
'the volumetric examination of the closure flange location will assure detection of the critical size flaw.
Volumetric examination of the reactor vessel flange-to-upper shell weld and specified adjacent base material is accomplished by two ultrasonic scan routines. Coverage from the flange side of the weld involves use of angled longitudincl. waves from the flange seal surface. Beam angles are selected based on their ability to provide coverage of the wf d and specified adjacent 1
base material to the extent practical and provide near normal incidence to the plane of the weld: Refracted beam angl'e's'in the range O to 16 are typically used for these examinations. Examinations from the shell side of the weld involve O*, 45', and 60 refracted angle beam coverage from the vessel inside-
. diameter surface. Angle beam scanning is performed in two directions, parallel to the weld and perpendicular to the weld from the shell side. Access for the shell side examination is limited to the Ten Year Inservice Inspection outage when the core barrel is removed from the reactor vessel.
The licensee indicates that the fact that postulated flaws are surface related is significant from a detection probability point of view.
Incipient cracks starting at right angles to a given surface (OD or ID) provide favorable conditions for detection via ASME Code specified 45" shear wave ultrasonic examinations from the opposite surface. Circumferential flaws are oriented favorably for detection during axial scanning. Axial flaws are oriented favorably for detection during circumferential scans. Circumferential1y oriented flaws in the vessel flange weld region also provide favorable conditions for detection during ultrasonic examinations from the flange seal surface.
Additional justifications for pemitting smaller postulated flaws in the closure flange region than the size postulated for the beltline region are described in the staff's regulatory analysis of public comments which is in to the staff's report SECY-83-80, "10 CFR Part 50-General Revision of Appendices G and H, Fracture Toughness and Reactor Vessel Material Surveillance Reouirements," February 25, 1983.
As previously reported, the licensee's fracture mechanics evaluation was performed in accordance with the methodology in Appendix A of ASME Code Section XI.
In this method, the stress intensity factors at the crack tip are calculated by linearizino the stress around the postulated flaw. The linearized stress is divided into membrane and bending stresses. The Appendix A method of linearizing stress resulted in negative membrane stresses
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4 when considering boltup, pressure and thermal condition during heat-up. The licensee considered the negative membrane stresses equal to zero when deter-mining the stress intensity factor resulting from thermal conditions during heat-up. The NRC staff considers this acceptable, since it conservatively l
represents the stress condition resulting from heat-up.
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The licensee used the negative value of membrane stress when determining the j
l stress intensity factor resulting from holtup and pressure conditions. The negative membrane stress will result in a reduction in the calculated stress intensity factor, since the stress intensity factor is the sum of a positive bending stress and a negative membrane stress. A negative value of membrane stress does not represent the real membrane stress resulting from boltup and pressure conditions. However, the non-conservatism resulting from a negative valued membrane stress will be offset by a high value for the bending stress that results from the linearizing method. Several methods of calculating l
stress intensity factors for a stress distribution similar to that in the closure flange region were evaluated in Reference 2.
The Appendix A method of linearizing the stress around the postulated flaw produced conservative stress intensity factors when compared to those calculated using a finite element analysismethod,anA'SMECodeSectionIII,AppendixGmethcdrecommendedfor nonlinear stress distributions, and a poly-nomial method (Reference 3). This comparison indicates that the Appendix A method of linearizing stress will result in an acceptable fracture mechanics analysis for evaluating flaws in the closure flange region of the reactor vessel.
Using the stress intensity factors calculated in accordance with Appendix A of the ASME Code Section XI and the safety margins of Appendix G of the ASME Code with a postulated flaw of 0.625 inch deep by 3.75 inches long, the licensee proposed pressure-temperature limits for the closure flange region materials.
The pressure-temperature limits for the closure flange region materials were combined with the limits for the beltline region to develop the Farley-1 Heatup/Cooldown Curves.
SAFETY
SUMMARY
Based on the method documented in Regulatory Guide 1.99, Revision 2 for evaluating the effect of neutron irradiation on reactor vessel beltline materials, and the method of calculating pressure-temperature limits in SRP 3.6.2, the licensee's proposed Heatup/Cooldown Curves for 16 EFPY meet the safety margins of Appendix G of the ASME Code.
Based on the licensee's finite element analysis, the fracture mechanics analysis performed in accordance with Appendix A of Section XI of the ASME Code, and the licensee's and NRC staff's justification for considering smaller postulated flaw sizes based on SECY-83-80, the licensee's proposed pressure-temperature limits for the closure flange region meet the safety margins of Appendix G of the ASME Code.
Based on the above two conclusions, the proposed Heatup/Cooldown Curves that are contained in the licensee's letter dated September 29, 1986, meet the safety margins of Appendix G, 10 CFR 50 for 16 EFPY and are acceptable Farley-1 Technical Specification.
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ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental l
assessment need be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded, based on the consideration discussed above, that: (1) there is reasonable a'ssurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the i
issuance of the amendments will not be inimical to the common defense and i
security or to the health and safety of the public.
References 1.
" Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems), U.S. Department of Comerce, December 1,1958 and February 27, 1959, pp. 58, 59, 60, Addendum No. 1.
2.
Bloom, J.M., and Van Der Sluys, W.A., " Determination of Stress Intensity Factors for Gradient Stress Fields," Journal of Pressure l
Vessel Technology, Vol. 99, August 1977.
3.
Buchalet, C.B., and Bamford, W.H., " Stress Intensity Factor Solutions for Continuous Surface Flaws in Reactor Pressure Vessels,"
Mechanics of Crack Growth. ASTM STP 590, American Society for Testing and Materials, 1976.
Principal Contributor:
B. Elliot Dated: June 23, 1987
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