ML19260D432
| ML19260D432 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/17/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17174A429 | List: |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR NUDOCS 8002110042 | |
| Download: ML19260D432 (3) | |
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?r SYSTEMATIC EVALUATION PROGRAM PLANT SYSTEMS / MATERIALS DRESDEN NUCLEAR POWER STATION UNIT NO. 2 Topic III-8.C - Irradiation Damage, Use of Sensitized Stainless Steel and-Fatigue Resistance The safety objective of this review is to detemine whether the integrity of the internal structures of operating reactors has been degraded through the use of sensitized stainless steel.
The effect of neutron irradiation and fatigue resistance on materials of the internal structures was eliminated from the safety objective of Topic III-8.C in memorandum to D. G. Eisenhut from D. K. Davis and V. S. Noonan dated Decemoer 8, 1978. The memorandum concluded that operating experience indicated that no significant degradation of the materials of the reactor internal structures had occurred as a result of either irradiation or fatigue.
Furthernare, the Standard Review Plan (Section 4.5.2) does not address neutron irradiation nor fatigue resistance of the materials of the reactor internal structures.
As a result i incidents of intergranular stress corrosion cracking in piping in the BWR c,ystem, special study groups were formed by NRC and industry to evaluate the cause, extent and safety implications of the use of sensitized stainless steel in the nuclear steam sucaly systems. The study groups identified tne incidents with the recirculation system bypass lines, the coro spray lines, and the contml rod drive return lines. It was concluded that the problem was caused by a combination of high total stresses, sens tization of the austenitic stainless steel in the heat affected :enes of welds, and the relatively high oxygen content of the coolant.
The NRC study group recom. ended an augmented inservice inspection program for stainless steel piping, more stringent monitoring of the leak detection system, modification of plant operating practice, and the use of alternate materials ix.une to intergranular stress corrosion cracking. The study grouo concluded in NUREG-0531, " Investigation and Evaluation of Stress-Corrosion Cracking in Picing of Lignt Water Reactor Plants," nat intergranular stress corrosion cracking in pioing would be detected prior to unstable crack growth because of the adequacy of the inservice inspection program and the leak detection system.
Reactor ocerating exoerience has validated the leak-before-break concept of piping integrity, and, it was concluded, that through-wall cracks in the piping systems would be detected before they presented a hazard to the health and safety of the public.
The regulatory position on the use of sensitized stainless steel in reactor interncl materials is addressed in the Standard Review Plan Section 4.5.2,
" Reactor Internal Materials." The areas currently reviewed in the applicant's SAR are materials specification and the controls imposed on the reactor coolant chemistry, fabrication practices and e:, amination and protecticn procedures.
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. The materials specification should comply with Section III of the ASME Boiler and Pressure Vessel Code and the components should satisfy the recommendations of Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal" and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel".
The reactor is described in Sections 3 and 4 of the Safety Analysis Report for the Dresden Nuclear Power Station Unit Nos. 2 and 3.
The internal components were designed to provide support for the fuel and maintain structural clea ances during normal and accident conditions.
In addition, the internal components provide passageway for the coolant to cool the fuel and means for adequately separating the steam from the coolant water.
The vessel was designed, fabricated and tested in accordance with Section III of the ASME Soiler and Pressure Vessel Code,1963 Edition, including Sumer 1964 Addenda. The materials used for the construction of the reactor internals were identified as Type 304 and Type 308 stainless steel, Inconel, and minor cuantities of soecial purpose alloys. The identified structural materials have been used on other General Electric designed reactors and have proven adequate by the results of extensive tests, prior usage and satisfactory performance.
The regulatory position on the use of sensitized stainless steel in reactor internal mate..-ials was not addressed in the Safety Analysis Report for the Dresden Nuclear Power Station Unit No 2.
Experience has shown that at least three elements in combira: ion are necessary to cause cracking in si.asitized stainless steel components. Tnese are material susceptibility, an oxygenated water environment, and a threshold total stres;. We assume for this evaluation that the Dresden Unit No. 2 reactor internal components contain sensitized stainless steel in contact with an oxygen saturated coolant water environment.
However, the calculated stresses on the reactor internal components do not exceed the threshold stress values associated with intergranular stress corrosion cracking. The threshold stress values are near or greater than the 0.2% off-set yield stress at temperature. Furtner, in the reactor environment, stress relaxation may occur due to trradiation and temperature effects.
The Licensee Event Reports and the BWR Nuclear Power Excerience were reviewed for the Dresden Nuclear Power Station Unit No. 2 in arcer to correlate reactor internal materials failure to the use of sensitized stainless steel in the components. The events are summarized.as follows:
In LER's beginning in September 1974, and continuing through 1977, leaks were reported in the 4-in recirculation bypass line, the 10-in core spray lines, the 14-in HPCI lines, and the control rod drive penetration. The cause of these failures was intergranular stress corrosion cracking, resulting from either furnace or weld sensitization. The cefective material was replaced with material immune to sensitization. An augTented inservice inspection program was conducted to reduce tne consecuence of the failure incidents in the arfected piping.
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,.. In December 1974, cracks in the feedwater spargers were discovered.
This problem is an item under review by Generic Technical Activity A-10, "BWR Feedwater Nozzle Cracking".
During inservice inspection of reactor internals in March 1976, loose restrainer clamp bolt keepers were found on the jet pumps.
The cause of the failure was fatigue cracking of the tack-welded restrainer assembly. The loose keepers were rewelded to the original specification.
We conclude from our review of the Licensee Event Reports and the BWR Nuclear Power Exoerience that the i.1tegrity' of the reactor internal components was cegracea oy tne use of sensitized stainless steel. The reported events were detected by the inservice inspection and testing program. The events were considered by the NRC Pipe Crack Study Group in NUREG-0531, " Investigation and Evaluation of Stress Corrosion Gradir.g in Piping of Light Water Reactor Plants" and in the Generic Technical Activity A-10, "BWR Feedwater Nozzle Cracking".
The inservice inspection program for the reactor internal components is being conducted during the current interval to the recuirements of Section XI of the ASME Soiler and Pressure Vessel Code,1974 5dition, including Sumer 1975 Addenda. The program is in compliance with paragraph (g) of Section 50.55a of 10 CFR Part 50. It will assure that the integrity of the comperents is maintained during reactor operation.
We conclude
- rom cur review of the information submitted by the licensee that the materials in the reactor internal components are sensitized and are coerated in an oxygen saturated water environment, and that the incidents of stress corrosion cracking are rare because the total stress level is relatively low, not exceeding the 0. '.' offset yield strength at operating temoerature.
In the unlikely event that intergranular stress corrosion cracking should occur, cracks in the comocnents will be detected by inservice inscection procedures crior to co ocnent failure We conclude that the intecrity of the reactor internal'c:moonents w1Il be assurec by the inservice lhspection program conducted to the recuirements of Section XI of the ASME Soiler and Pressure Vessel Code,1974 5dition, including Sumer 1975 Addenda, in comoliance with. paragraph (g) of Section 50.55a of 10 CFR Part 50. Further, we conclude that intergranular stress corrosion cracking in the reactor internal compenents is not a hazard to the health and safety of the public.
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