ML20138C465
| ML20138C465 | |
| Person / Time | |
|---|---|
| Site: | Bellefonte  |
| Issue date: | 11/20/1985 |
| From: | Hufham J TENNESSEE VALLEY AUTHORITY |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8512130025 | |
| Download: ML20138C465 (2) | |
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TENNESSEE VALLEY AUTHORITY CH ATTANOOGA. TENN SEE 37401 SN 157B Looko
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November 20, 1985 BLRD-50-438/83-43
.BLRD-50-439/83-36 U.S. Nuclear Regulatory Commission Region II Attn:
Dr. J. Nelson Crace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Dear Dr. Grace:
BELLEFONTE NUCLEAR PLANT UNITS 1 AND 2 - REACTOR BUILDING SUMP PH VALUES BY BABCOCK AND WILC0K - BLRD-50-438/83-43 AND BLRD-50-439/83 FINAL REPORT The subject deficiency was initially reported to NRC-0IE Inspector Linda Watson on June 23, 1983 in accordance with 10 CFR 50.55(e) as NCR BLN NEB 8307. This was followed by our interim reports dated July 21, 1983 and January 12 and July 16, 1984.
Enclosed is our final report which concludes that we no longer consider this deficiency to be reportable under the provisions of 10 CFR 50.55(e).
If you have any questions, please get in touch with R. H. Shell at FTS 858-2688.
Very truly yours, TENNESSEE VALLEY AUTHORITY J
W.
fjam, anager L:. ensing and Lsk Protection Enclosure cc:
Mr. James Taylor, Director (Enclosure)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Comnission Washington, D.C.
20555 Records Center (Enclosure)
Institute of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339 H. B. Barkley, Manager (Enclosure) 205 Plant Project Services P.O. Box 10935 Lynchburg, Virginia 24506 8512130025 851120 3
PDR ADOCK 05000438 g
PDR An Equal Opportunity Employer
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 5N 157B LookoqtjPg g b
November 20, 1985 BLRD-50-438/83-43 BLRD-50-439/83-36 U.S. Nuclear Regulatory Commission Region II Attn:
Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Dear Dr. Grace:
BELLEFONTE NUCLEAR PLANT UNITS 1 AND 2 - REACTOR BUILDING SUMP PH VALUES BY BABCOCK AND WILCOX - BLRD-50-438/83-43 AND BLRD-50-439/83 FINAL REPORT The subject deficiency was initially reported to NRC-OIE Inspector Linda Watson on June 23, 1983 in accordance with 10 CFR 50.55(e) as NCR BLN NEB 8307. This was followed by our interim reports dated July 21, 1983 and January 12 and July 16, 1984. Enclosed is our final report which concludes that we no longer consider this deficiency to be reportable under the provisions of 10 CFR 50.55(e).
If you have any questions, please get in touch with R. H. Shell at FTS 858-2688.
Very truly yours, TENNESSEE VALLEY AUTHORITY h>
J W. H fjam, Manager L:. ensing and R'Lsk Protection Enclosure cc:
Mr. James Taylor Director (Enclosure)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Records Center (Enclosure)
Institute of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339 H. B. Barkley, Manager (Enclosure) 205 Plant Project Services P.O. Box 10935 Lynchburg, Virginia 24506 8S12130025 851120 PDR ADOCK 05000438 S
PDR An Equal Opportunity Employer y6M t
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ENCLOSURE BELLEFONTE NUCLEAR PLANT UNITS 1 AND 2 REACTOR BUILDING SUMP PH VALUES BY' BABCOCK AND WILCOI BLRD-50,-438/83-43 AND BLRD-50-439/83,
NCR BLN NEB 8307 10 CFR 50.55(e)
FINAL REPORT Description of Deficiency In response to NRC's FSAR question 281.2, TVA requested Babcock and Wilcox (B&W) to recalculate reactor building (RB) sump pH values as necessary to include consideration of " dead" volumes which are not available for recirculation. The major contribution to the total " dead" volume is the reactor vessel (RV) cavity (55 percent of total volume).
In performing the
-recalculation, B&W has found some problem areas.
Depending on where the break is assumed..the RB pH may exceed the limits set by the NRC (8.5 ( pH 4[11.0)
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for prevention of stress corrosion and for materisis compatibility.
Safety Implications TVA has analyzed the effects of the revised pH parameters and has adequate documentation-to show that the sump pH being outside the recommended limits set by the NRC is not detrimental to plant safety in either the short-term or long-term. As the materials in the " dead" volumes are limited in type and function, the scope of the analysis was also limited. Basically, three material concerns were addressed:
(1) the effects on strength and chloride leachability of concrete samples exposed to the conditions anticipated in the RV cavity.
(2) the effect of galvanic polarization corrosion of dissimilar metals of the incore instrument tubes in the low pH water in the reactor cavity.
(3) the effect of stress corrosion cracking on the type 304L stainless steel.
incore monitor lines.
The compressive strength of the concrete specimens subjected to the borated reactor coolant solution at 100 C was essentially equivalent to that of 0
concrete exposed to air at a-temperature of 200 C.
0 Leachability tests indicate that relatively little chloride will be added to the bulk solution due to leaching of the concrete since the solution pH increases to approach a value of 7 with a corresponding slowdown in leaching rate.
Galvanized polarization corrosion of dissimilar metals was investigated and the corrosion rates were determined to be acceptable.
Stress corrosion cracking of 304L stainless steel welds was investigated and determined to not occur at the conditions expected af ter 'a loss of coolant accident (LOCA).
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In. completing the work necessary for the analysis above, it was determined that the sodium hydroxide (NaOH) concentration in the NaOH tank should be reduced from 22 2 to 18 2 wt%. The reduction in NaOH concentration assists in minimizing,th,e higher pH levels where NaOH injection is initla-ted as part of the emergency core cooling system (ECCS).
It should be noted that this reduction in NaOH toncentration is not required due to safety ~~
considerations but is a change due to economic considerations as a result of the more refined analysis.
TVA also considered a small break loss of coolant accident (SBLOCA) where:
(a) building spray with NaOH addition is not initiated automatically.
Break location is outside of reactor vessel cavity and at top of loop -- reactor coolant system remains filled or is refilled.
(b) After sump recirculation has begun, operator initiates NaOH addition to the. low pressure injection (LPI) system.
In this mode of operation, the sump pH could be administratively controlled by the operator according to a plot of NaOH tank level versus pH.
It is theoretically possible to reach a sump pH of 12 if the reactor coolant contains O ppm Boron and the NaOH concentration is at its maximum (20%).
However, based on the low probability of an SBLOCA, no operator action until recirculation, all conditions being at extreme limits assumed in the calcula-tion and conservatisms used to perform calculations, we do not feel this is a significant concern.
Consequently, TVA no longer considers this deficiency to be reportable under the provisions of 10 CFR 50.55(e).
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