ML20036A492
| ML20036A492 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/04/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20036A477 | List: |
| References | |
| NUDOCS 9305110336 | |
| Download: ML20036A492 (4) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GEORGIA POWER COMPANY. ET AL.
EDWIN 1. HATCH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By [[letter::HL-2006, Application for Amends to Licenses DPR-57 & NPF-5,revising TS to Remove Main Steam Line Radiation Monitor Reactor Scram & Group Isolation Functions,Per BWROG Topical Rept NEDO-31400, SE for Eliminating BWR MSIV Closure..|letter dated October 19, 1992]], Georgia Power Company, et al. (GPC or the licensee), proposed to change the schedule for the liquid penetrant (PT) examinations and the ultrasonic tests (UT) contained in NUREG-0619 "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking."
The licensee proposed to implement the proposed change during the current spring 1993 refueling outage for Unit I and the spring 1994 refueling outage for Unit 2.
By letter dated January 25, 1993, the NRC staff requested additional information to evaluate the acceptability of the proposed alternative inspection.
By letter dated February 2,1993, the licensee limited its.
original request, at this time, to a schedule change for the PT examination for one cycle only for both units.
Furthermore, by NRC 930125 RAI on NUREG-0619 Insp Requirements,As Modified by GL 81-11.Rept Withheld,Per 10CFR2.790|letter dated February 22, 1993]], GPC provided its response to NRC's request for additional information.
This safety evaluation (SE) addresses only a one-time, one fuel cycle deferral of the scheduled PT examination at both units. The licensee had performed UT inspections at both units every second refueling outage as specified in Table 2 of NUREG-0619. No reportable indications have been detected.
In addition, a limited scope PT of two feedwater nozzles was performed in 1985 at Unit 2.
2.0 BACKGROUND
The NRC staff published NUREG-0619 as the resolution of Generic Activity A-10 on BWR feedwater nozzle cracking. The NUREG was modified by Generic Letter (GL) 81-11 dated February 20, 1981. The above documents addressed a combination of potential solutions to the feedwater nozzle cracking problem which includes removal of the stainless steel cladding, acceptable thermal sleeve designs, modification of plant systems, operating procedural changes, bypass leakage detection systems, and periodic inservice inspections.
NUREG-0619 did not centain rscommendations for a permanent solution but provided an interim guidance for plant-specific action.
The licensee committed to comply with the inspection schedule in Section 4.3.2, Table 2, of NUREG-0619 in a letter dated January 22, 1981 For the existing feedwater nozzle configurations at both Hatch units, Table 2 of NUREG-0619 specifies an external UT examination of all feedwater nozzle
$!k i
. safe-ends, bores, and inside blend radii every other refueling outage.. If the routine UT detects recordable indications that are interpreted to be cracks in any nozzle, the sparger must be removed, and examined with PT.
If any cracks are detected, all spargers must be removed and completely examined and all cracks must be removed.
For a welded thermal sleeve design, similar to Hatch Unit 2, Table 2 specifies a routine PT of the accessible areas after six refueling outages cr 90 startup/ shutdown cycles.
For a feedwater nozzle with a triple-riceve sparger, similar to Unit 1, Table 2 specifies a routine PT of one nozzle after nine refueling outages or 135 startup/ shutdown cycles. Table 2 of NUREG-0619 and GL 81-11 state that the routine PT must be performed even if the scheduled UT and leak test results are satisfactory.
Section 4.3.2.3 of NUREG-0619 specifies that all cracks must be removed.
3.0 LICENSEE'S TECHNICAL BASES GPC's [[letter::HL-2006, Application for Amends to Licenses DPR-57 & NPF-5,revising TS to Remove Main Steam Line Radiation Monitor Reactor Scram & Group Isolation Functions,Per BWROG Topical Rept NEDO-31400, SE for Eliminating BWR MSIV Closure..|letter dated October 19, 1992]], stated that a reliable technique is now available to ultrasonically inspect the feedwater nozzle inner radius and the bore region.
This technique proved to be capable of detecting and sizing flaws 0.25 inches or less in depth. Furthermore, the licensee stated that since NUREG-0619 was issued, no incidents of nozzle bore or inner blend radius cracking have been reported at plants where vessel cladding in the nozzle area was removed.
This fact, combined with the installation of improved design spargers that minimize bypass flow leakage, appears to have significantly reduced, if not eliminated, the feedwater nozzle cracking problem. Therefore, GPC concluded that the nozzle issue has been adequately resolved through implementation of NUREG-0619 recommended actions and UT examination in the area of interest by:
1.
Eliminating the different thermal expansion rates between the nozzle cladding and the vessel, 2.
Reducing / eliminating the crack initiating leakage by installing improved
- spargers, 3.
Improving system operations to minimize thermal cycling, and 4.
Showing by analysis that an assumed crack of 0.25 inch would not grow to exceed the allowable 1.00-inch depth. The fracture mechanics analysis for Unit 2 was accepted by the staff and the SE was transmitted by letter dated January 30, 1992.
The licensee described the burden and potential radiation exposure associated with performance of the routine PT recommended by the NUREG.
Based on the 8 person-rem cumulative dose received by inspectors during the 1985 limited scope PT of two feedwater nozzles at Unit 2, the licensee estimated that a dose of at least 25 person-rem would be accumulated by performing a routine PT examination which involves the removal of a sparger in Unit 1.
Consequently, GPC requested a schedule change for the PT examination recommendation contained in NUREG-0619 for one cycle on both Units 1 and 2.
' 4.0 STAFF EVALUATION The licensee proposed to perform an automated UT that is capable of sizing and detecting small thermal fatigue cracks, as a part of the basis for deferral of the routine PT for one fuel cycle. The staff agrees that automated UT techniques have improved since 1981 when the NUREG was published and that certain advanced UT, with computer-based data acquisition systems and programmed scanners, are capable of detecting a pattern of thermal fatigue cracks in the nozzle inner radius and bore region.
In addition, manual UT techniques, developed for the detection and sizing of IGSCC in stainless steel piping, are applicable for the detection of thermal fatigue cracking in the nozzle safe-end region.
However, UT from the outside diameter (OD) surface of the reactor vessel will not detect the small flaws in the inner radius, bore and safe-end region that the PT recommended by NUREG-0619 would detect.
To reach a determination regarding one fuel cycle deferral of the PT, the NRC staff did not evaluate whether a 1/4-inch deep flaw could be reliability detected in the inner radius.
For this review, the NRC staff used the assumed 1.00-inch deep end-of-life crack acceptance criteria described in NUREG-0619 and GL 81-11. Thus, a deferral of the PT is acceptable provided that the automated UT will reliably and repeatedly detect a pattern of thermal fatigue cracks. However, if these cracks are missed, they could propagate to a depth of only 1.0 inch between inspections.
The licensee stated that they will use the General Electric Remote Inspection System (GERIS) for the UT of the feedwater nozzle inner radius and bore areas in Unit I during the spring 1993 refueling outage. The safe-end examinations will be performed using the General Electric " SMART 2000" automated UT piping inspection system. The three welds on each of the Unit 2 safe-ends are included in the NUREG-0313 and GL 88-01 inspection scope.
Based on the IGSCC examinations (i.e., that the detection capabilities for the safe-end region have been established), the NRC staff finds that_ the deferral is acceptable for one cycle.
By letter dated January 30, 1992, the NRC staff transmitted its SE on Unit 2 feedwater fracture mechanics analysis which discussed the number of events, cycles, and thermal conditions projected to occur during 40 years that could contribute to thermal fatigue in the nozzle inner radius region. A 0.25-inch deep flaw was assuned in the licensee's analysis. This conservatively enveloped the initial crack that was assumed in NUREG-0619 that could be left undetected during repair / original construction. The NRC staff concluded that after 40 years of operation, based on the actual measurements at Hatch Unit-2, the final crack depth would be less than 1.00 inch. A similar fracture mechanics was performed for Hatch Unit 1.
To confirm that unacceptable flaw growth has not occurred, the licensee had performed ultrasonic testing every second refueling at both units. No reportable indications have been detected.
In Mdition, limited scope interval PT was performed on Unit 2 in 1985.
In response to the staff's request for additional information, the licensee provided a General Electric
. proprietary report (a non-proprietary version will be provided in the future) describing the detection capabilities of the UT usi'g GERIS and a summary of n
all inspections performed at Plant Hatch since 1977. The last UT was performed on all feedwater nozzles in 1991 and 1992 at Units 1 and 2, respectively.
Based on the NRC staff's review of the licensee's submittal, the staff concludes that an automated ultrasonic test will detect a pattern of significant thermal fatigue cracks in the nozzle inner radius region that is consistent with the plant-specific fracture mechanics analysis reviewed pursuant to GL 81-11. Therefore, the staff concludes that a one-cycle schedular change for both units from the PT examination recommendation is acceptable provided that the licensee implement its plan outlined in the October 19, 1992, submittal, subject to the following conditions:
1.
Perform an automated ultrasonic test of all nozzle inner radius, bore and safe-end regions at Hatch Unit I with the GERIS and SMART 2000 system.
If a flaw indication is detected that is interpreted to be a crack, the corrective action PT described in NUREG-0619 must be implemented.
2.
Submit the technical bases and sufficient information to address the acceptability of the spring 1993 examination results before February 1, 1994.
3.
Perform an automated ultrasonic test at Hatch Unit 2 with a similar inspection system and examination procedure as that of Unit 1.
The licensee should incorporate any improvements or revisions from item 1 above.
4.
Submit the technical bases and sufficient information to address the acceptability of the spring 1994 inspection at Hatch 2 before September 1, 1994.
Based on the plan described above, the staff will resolve the issue of i
appropriate inspection methods and frequencies before the next Hatch Unit I refueling outage currently scheduled for the fall of 1994. Based on the information described above, the NRC staff concludes that an automated UT of all feedwater nozzles will provide an acceptable level of quality and safety.
Principal Contributor:
M. Hum, NRR Date:
,