ML20203N620
| ML20203N620 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/21/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203N591 | List: |
| References | |
| TAC-59902, NUDOCS 8605050555 | |
| Download: ML20203N620 (5) | |
Text
l UNITED STATES 8
NUCLEAR REGULATORY COMMISSION E
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO. 55 TO FACILITY OPERATING LICENSE NO. NPF-8 l
ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT. UNIT NO. 2 DOCKET NO. 50-364 Introduction In a letter from R. P. Mcdonald to S. A. Varga dated September 30, 1985, Alabama Power Company (the licensee, APCo) requested changes to the Joseph M.
Farley Nuclear Plant, Unit 2 heatup/cooldown curves and supporting bases.
The curves and bases are contained in Figures 3.4-2 and 3.4-3 and Bases Section 3/4.4.10 of the Farley 2 Technical Specifications.
The effect of neutron irradiation on the Farley 2 reactor vessel beltline materials is documented in Westi6ghouse Report WCAP-10910, which is enclosed in the licensee's letter of September 30, 1985. The effect of boltup, pressure and thermal stresses on the i
reactor vessel closure flange region is documented in Attachment 2 to the licensee's letter of September 30, 1985 and in a previous letter from R. P. Mcdonald to S. A. Varga dated June 18, 1984.
By letter dated March 27, 1986, APCo provided supp1mntary information following discussions with the l
1:RC staff. Our discussion and evaluation follows.
-Discussion and Evaluation Heatup/cooldown curves must be calculated in accordance with the requirements of Appendix G, 10 CFR 50, which became effective on July 26, 1983.
Appendix G, 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 with these requirements, the effect of neutron irradiation, boltup, pressure and thermal stresses on the
. limiting reactor vessel beltline and closure flange region materials must be estimated.
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The method recommended by the NRC staff for calculating the effect of neutron i
irradiation damage is documented in Regulatory Guide 1.99, Rev. 1, " Effects i
of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."
This guide indicates that when credible surveillance data becomes available, 1
increases in reference temperatures resulting from neutron irradiation damage may be predicted by extrapolating the surveillance data to higher or lower i
fluences following the slope of the family of curves in Figure 1 of the guide.
The limiting meterial in the Farley 2 reactor vessel beltline is Plate B7212-1.
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Samples from this plate were placed in the Farley 2 reactor vessel surveillance capsules for irradiation and testing.
Test results on this irradiated plate material were reported in Westinghouse Report WCAP-10425, " Analysis of Capsule U from the Alabama Power Company, Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program" in APCo letter dated November 10, 1983.
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,, 1 Since material from the limiting plate has been placed in the Farley 2 surveillance capsules, test results after irradiation of the capsules will produce credible surveillance data.
Hence, we have evaluated the effect of neutron irradiation on the Farley 2 reactor vessel by extrapolating the surveillance test data from samples of Plate B7212-1.
This extrapolation was done in accordance with Regulatory Guide 1.99, Rev. 1.
Using the curvesmeetthesafetymarbs,thelicensee'sproposedheatupandcooldown extrapolated values for RT ataneutronfluenceof7.7x10{8ppegdixGoftheASMECodeSectionIII o A r1/cm (E>1MeV).
This corresponds to eight effectise full power years (EFPY) of operation.
The licensee's subnittal of September 30, 1985, includes a bases section and figures showing the proposed pressure / temperature limits.
The pressure / temperature limits are proposed for nine EFPY.
But, our review shows that the licensee's proposed calculation method does not include sufficient margin to account for neutron irradiation damage.
Hence, the proposed bases section required revision.
Also, the proposed heatup/cooldawn curves would meet the safety margins of Appendix G of the ASME Code Section III for only 8.0 EFPY instead of the proposed 9.0 EFPY.
For these reasons, we advised the licensee and Westinghouse of our evaluation during various telecons.
As a result, by letter dated March 27, 1986, the licensee modified their earlier bases section and provided heatup/cooldown curves for 8.0 EFPY which are acceptable.
The stresses resulting from pressure, thermal and boltup on the closure flange region 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 four loop reactor vessel.
However, the Farley 2 plant contains a three loop reactor vessel.
The geometry of-the closure flange region in Farley 2 reactor vessel is slightly different than that of the typical four loop reactor vessel.
To account for these differences, the licensee performed a dimensional stress analysis of the two types of vessels.
Their analysis indicates that the typical four loop reactor vessel and the Farley 2 reactor vessel have essentially equivalent stresses resulting from pressure and boltup in the critical closure flange region.
Hence, the stresses from boltup and pressure used for the typical four loop plant were used in the fracture mechanics evaluation for Farley 2.
The stresses resulting from thermal conditions during heatup or cg1down of the Farley 2 vessel were determined by the computation method - to be significantly less than those calculated for the typical four loop plant.
The thermal stresses at the critical closure flange region in the Farley 2 reactor vessel were calculated yj by reducing the finite element thermal stresses for the typical four loop reactor vessel by the ratio of the thermal stresses in the three loop to those in the four loop.
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" Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems),"
U.S. Department of Commerce, December 1, 1958 and February 27, 1959, pp. 58, 59, 60, Addendum No. 1.
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3 Fracture mechanics evaluation at three discontinuity locations in the closure flange region were performed in accordance with the methodology in Appendix A of ASME Code Section XI.
In this analysis the licensee used all the safety factors required by Appendix G of the ASME Code, except for the Code recommended flaw size, to determine the closure flange location that would be 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 determined to be the outside surface at the discontinuity between the flange and shell of the reactor vessel.
The postulated flaw size recommended by Appendix G of the ASME Code was used for evaluating the beltline region, but was not used in evaluating the closure flange region.
The postulated flaw has a depth of 1/4 the section thickness (1/4 T) and a length of 1 1/2 times the section thickness.
The section thickness at the critical flange location for Farley 2 is 9.125 inches.
Appendix G of the ASME Code indicates that small defect sizes may be used on an individual case basis, if a smaller size of maximum postulated defect can be assured.
The postulated defect used in the licensee's analysis was a 0:625 inch deep by 3.75 inches long 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 volumetric examination of the closure flange location will detect this 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 longitudinal waves from the flange seal surface.
Beam angles are selected based on their ability to provide coverage of the weld and specified adjacent base material to the extent practical and provide near normal incidence to the plane of the weld.
Refracted bea:n angles in the range 0*
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 examinations is limited to the ten year ISI outage when the core barrel is removed from the reactor vessel.
The licensee indicates that the fact that the 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.
Circumfer-entially oriented flaws in the vessel flange weld region also provide favorable conditions for detection during ultrasonic examinations from the flange seal surface.
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- Additional justifications for permitting smaller postulated flaws in the closure flange region than that postulated for the beltline region are described in Enclosure 4 to the NRC staff's report SECY-83-80, "10 CFR Part 50 -
General Revision of Appendices G and H, Fracture Toughness and Reactor Vesssi Material Surveillance Requirements," February 25, 1983.
As previously reperted, 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 linearizing the stress around the postulated flaw.
The linearized stress is divided into membrane and bending stresses.
The Appendix A, ASME Code Section XI method of linearizing stress resulted in negative membrane stresses when considering boltup, pressure and thermal condition during heat-up.
The licensee considered the negative membrane stresses equal to zero when determining the stress intensity factor rcsulting from thermal conditions during heat-up.
The staff considers this acceptable, since it conservatively represents the stress condition resulting from heatup.
The licensee used the negative value of membrane stress when determining the stress intensity factor resulting from boltup 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 may be offset by a high value for the bending stress that results from the linearizing method.
The NRC staff has not completed its evaluation of this issue.
The NRC staff is discussing this issue with individuals who are members of the ASME Code Subcommittee on Flaw Evaluation.
If we determine that the use of negative-valued membrane stresses and high bending stresses calculated in accordance with the Appendix A, ASME Code Section XI method of linearizing stresses results in non-conservative stress intensity factors, we will supplement this evaluation and inform the licensee that the approved pressure-temperature limits rbay require further adjustment.
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 mpterial were incorporated into the proposed Farley 2 heatup/cooldown curves.
Safety Summary 1)
Based on the method documented in Regulatory Guide 1.99, Rev.1 used for evaluating the Farley 2 surveillance data and reactor vessel beltline materials, the licensee's proposed heatup/cooldown curves will meet the safety margins of Appendix G of the ASME Code for 8 EFPY.
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- e. 2)
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 our further justification for considering smaller postulated flaw sizes (based on SECY-83-80), the licensee's proposed pressure-temperature limits for the closure flange region will meet the safety margins of Appendix G of the ASME Code.
Based on our review and on the above two conclusions, we conclude that the modified heatup/cooldown curves provided in licensee letter dated March 27, 1986, meet the safety margins of Appendix G,10 CFR 50 and are acceptaMe for eight EFPV of operation.
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 ne 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.
Accordir. gly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance 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 issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
April 21, 1986 Principal Contributor:
B. Elliot
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