ML18057A959
| ML18057A959 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/30/1991 |
| From: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9106110394 | |
| Download: ML18057A959 (3) | |
Text
-f consumers Power POW ERi Nii MICHlliAN'S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 May 30, 1991 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
GB Slade General Manager PERFORMANCE OF A REDUCED PRESSURE CONTAINMENT INTEGRATED LEAK RATE TEST (ILRT)
FOLLOWING STEAM GENERATOR REPLACEMENT The purpose of this letter is (1) to document our current understanding of the NRC Region III interpretation of 10 CFR 50 Appendix J regarding reduced pressure containment integrated leak rate testing following steam generator replacement and (2) to provide the NRC with our interpretation of what we believe to be the applicable portions of 10 CFR 50 Appendix J regarding this same issue.
On February 11, 1991, an entrance meeting was conducted at Palisades to discuss the NRC Region III plans to witness the structural integrity test (SIT) and the containment integrated leak rate test (CILRT).
During the entrance meeting the NRC Region III inspector informed Palisades Plant management of the Region III interpretation concerning the performance of a reduced pressure containment integrated leak rate test (ILRT).
The Region III staff indicated that since Palisades has opened and closed a large construction opening in the containment building, pre-operational ILRT test data must be re-established.. In other words, prior to continuing the practice of performing a reduced pressure containment ILRT, Palisades should perform a reduced pressure containment ILRT followed by a full pressure containment ILRT to establish the maximum allowed leak rate at the reduced pressure.
The Region III interpretation is based on 10 CFR 50 Appendix J, Section III.4(a)(l) entitled "Pre-operational Leakage Rate Test."
In resp~nse to the Region III interpretation, we believe that 10 CFR 50 Appendix J, Section IV, "Special Testing Requirements" applies for the modification performed on the containment building that allowed removal and installation of the steam generators; Under Subsection A, "Containment Modificatiqns," it is stated that, "Any major modification, replacement of a component which iscpart of the primary reactor containment boundary, or resealing a seal welded door, performed after the pre-operational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable for the area affected by the modification "(emphasis added).
We conclude that the Appendix J authors recognized that modifications would be A CMS ENERGY COMPANY
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made after performance of the pre-operational leakage test and that re-performing the pre-operational leakage test was not necessary.
During a conference call with Region III staff, NRR-Systems staff and the Palisades NRC Project Manager held April 26, 1991, NRR staff stated that, since the plant was already operational, technically, the test could not be called a pre-operational test, in accordance with Appendix J,Section III.A.4 (a)(l). However, NRR staff indicated that it would be a "good idea" to re-establish the correlation between the reduced pressure test and the full pressure test (defined in Appendix J,Section III.A.4(a)(l)(iii)), because we had opened and closed the containment building wall for steam generator replacement.
We believe that the significant leakage rate contributors are those valves and penetrations for which 10 CFR 50, Appendix J, Type B and Type C test are required at more frequent intervals than the Type A containment ILRT.
Furthermore, we believe that the replacement of the section of containment wall and liner plate, which serves as the leak tight barrier, is of minor significance to the containment integrated leakage rate. The liner plate replacement included non-destructive examination of the liner plate welds and a complete vacuum box test to show weld integrity. Therefore, the contribution of any leakage through the liner plate during an integrated leak rate test or actual accident event would be extremely minor.
We conclude that performing a Type A (ILRT) test following the restoration of the containment building meets the requirements of 10 CFR 50 Appendix J.
We also point out that 10 CFR 50 Appendix J, Section IV.A, would have allowed a reduced pressure ILRT to be performed following the containment modifications.
We voluntarily applied for an amendment to the Palisades Technical Specifications to allow a full pressure test to be conducted following steam generator replacement.
The Palisades Technical Specifications had restricted the ILRT pressure to a maximum of about 28 psig.
Had we not requested a change to the Technical Specifications, a reduced pressure test would have been conducted in accordance with both Appendix J and the Technical Specifications.
We are fully cognizant of the NRC efforts to revise 10 CFR 50, Appendix J, requirements to eliminate reduced pressure testing. Furthermore, several discussions have been held in the past between NRC ~taff and Palisades staff on the conduct of reduced pressure testing.
We believe that the reduced pressure ILRT, as allowed by 10 CFR 50, Appendix J, is a justifiable and an adequate surveillance of the containment leakage rate.
Based on our review of the time required to perform the full pressure containment ILRT versus the reduced pressure containment ILRT, we estimate that performing the full pressure containment ILRT will add approximately 1.5 days to a typical outage critical path.
Another issue which arose during our review of the inspection report 91003 was whether this containment ILRT test should be counted as the third test in ten years for the second ten year service period, as defined in Appendix J,Section III.D.l.
Based on our review of the methodology employed by Palisades to calculate (1) the as-found leak rate of containment prior to cutting the steam generator replacement opening and (2) the as-left leak rate calculated during the ILRT, we have concluded that this test will be counted as the third test in ten years.
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- In conclusion, we beli~ve that the pre-operational testing requirements of 10 CFR 50, Appendix J, Section III.4(a)(l) are applicable.
We believe that the Type A test, conducted during this past refueling outage, was done in accordance with 10 CFR 50, Appendix J, Section IV.A., "Containment, Modification," and no further testing is required to establish the maximum allowed leak rate at a reduced pressure beyond that originally established during the original pre-operational testing for the Palisades containment building.
Further, we believe that performance of a pre-operational test would not serve a useful purpose due to the minor effect that the containment structure modification could have on the containment integrated leak rate.
Therefore, Palisades plans to continue conducting containment ILRT's in accordance with Appendix J requirements which presently allow reduced pressure testing. If further discussion of this issue is necessary, please contact Cris Hillman at (616) 764-8913, extension 2173.
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Gerald B Slade General Manager CC Administrator, Region III, USNRC Resident Inspector, Palisades