ML19262C151
| ML19262C151 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/17/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19262C143 | List: |
| References | |
| NUDOCS 8002070021 | |
| Download: ML19262C151 (3) | |
Text
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SYSTEMATIC EVALUATION PROG:.A
PLANT SYSTEMS / MATERIALS DRESDEN NUCLEAR POWER STATIO" U:IT NO. 1 Topic III-8.C - Irradiation Dar. age, Use of Sensitized Stainless Steel and Fatigue Resistance The safety objective of this review is to determina whether the integrity of the internal structures of operating rectors has been degraded through the use of sensitized stainless steel.
Tne effect of neutron irradiation and fatigue resistance on materials of the internal structures was eliminated from the safety cojective of Topic III-8.C.
in memorandum to D. G. Eisenhut from D. K. Davis and V. S. Noonan dated December 8, 1979. The memorandum concluded that c:erating experience indicated that no significant degradation of the materials cf the reactor internal structures had occurred as a result of either irradiation or fatigue.
Furthen cre, 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 of incidents of intergranular stress corrosion cracking in piping in the SWR system, special study groups were fomed by THC and industry to evaluate the cause, extent and safety implications of the use of sensitized stainless steel in the nuclear steam supply syste s.
The study gi.ups identified the incidents with the recirculation system bypass lines, the core spray lines, and the control rod drive return lines.
It was concluded that the problem was caused by a combination of high tctal stresses, sensitization of the austenitic stainless steel in the heat affe::ed zones of welds, and the relatively high oxygen content of the coolant.
The NRC study group recorrendad an augmented inservice insoection progran for stainless steel piping, more stringent monitoring of the leak detection system, nooification of plant operating practica, and the use of alternate materiais immune to intergranular stress corrosion cracking.
The study group concluded in NUREG-0531, " Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants," that intergranular stress corrosion cracking in piping would be detected prior to unstable crack growth because of the acequacy of the inservice inspection program and the leak detection system.
Reactor coerating experience has validated the leak-before-break concept of pipin; integrity, and, it was concluded, that throuch-wall cracks in the piping systems would be detected before they presented a hazard to the health and sa#ety cf the publi 9
0 The reg.;1 story position on the use cf sensitized stainless steel in reactor internal rateriais is addressed in the Standard Review Plan Section 4.5.2,
" React:- Internal Materiais." The areas currenti;. reviewed in the applicant's SAR are materials specification and the controls ir. posed on the reactor coolant chemistry, fabrication practices and examination and protection procedures.
The ma erials specification snould comply with Section III of the ASME Boiler and Pressure Vessel Code and the components shoulc satisfy the recommenoations of Regula ory Guice 1.31, " Control of Ferrite Content in Stainless Steel Weld 8002070 2
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- Metal" and Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel."
The reactor vessel for the Dresden Nuclear Power Station Unit No. I was de-signed, fabricated and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section I and Section VIII, 1955 Edition, including Summer 1957 Addenda and applicable Code Cases. Stresses were calculated in accordance with paragraph UCL-23(b) of Section VIII of the Code.
The reactor internal structure is described in the Technical Section of the Final Hazards Summary Report for the Dresden Nuclear Power Station Unit No, l.
The internal components were designed to provide support for the fuel and maintain the required configuration and clearances during normal and accident conditions.
In addition, the internal components provide passageways for the coolant to cool the fuel and means for acequately separating the steam frc= the coolant water. The primary criteria for material selection for the reactor internal components were the mechanical procerties, the material stability and ccrrosion resistance in the reactor environment.
The materials used for fabricating the reactor internal components were identified in the Final Hazards Summary Report as Types 304 and 405 stainless steel, Incoloy, and minor quantities of soecial purpose alloys, such as 17-4 PH alloy.
Experience has shown that at least three elemerts in cc-bination are necessary to cause cracking in sensitized stainless steel comocnents.
These are material susceptibility, an oxygenated water environment, and a threshold total stress. We assume for this evaluation that the Dresden Nucl+ar Power Station Unit No. 1 reactor internal components contain sensitized stainless steel in contact with an oxygen saturated coolant water environment.
- However, the calculated stresses on the reactor components do not exceed the threshold stress values generally associated with intergraralar stress corrosisn cracking.
The threshold stress values are near or greater than tne 0.2% off-set yield stress at temperature.
Further, in the reactor enviror ent, stress relaxPtion may occur due to irradiation and temperature effects.
The Licensee Event ReDorts and the BWR Nuclear Power Experience were reviewed for the Dresden Unit No. 1 in order to correlate reactor internal materials failure and the use of sensitized stainless steel in the components.
The events are sum.arized as follows:
In November,199, the inservice inspection program revealed seDaration of the drive from the blade in one of the control rod units.
The failure was traced to intergranular stress cerrosion cracking of the 17-4 PH material aged at 9000F, This material was removed from all the control rod units and replaced with 17-4 PH material aged at 11000F, Test showed the lauer material not susceptible to intergranular stress corrosion.
1926 011 The inservice inspection program conductec in Feburary, 1951, revealed cracks in the blades of the control rod units.
The blades were boron stainless steel.
The cause of cracking was dae to the extreme brittleness of the material.
The blac;es b all the control rod units were replaced with boron carbide encapsulated in stainless steel tubes.
The latter design proved to be satisfactory for operating reactors.
The core support structure was fabricated from Type 405 ferrite stainless steel.
During welding, martensite may form in the heat-affected-zone, resulting in embrittlement.
Inspection of the grid structure during fabrication revealed
" toe" cracks, which were repaired.
At the first inservice inspection, eight additional cracks were observed.
It was concluded that these cracks originated during fabrication and did not impair the strength of the core support structure.
Subsequent inservice inspections showed that there was neither extension nor worsening of existing defects, any new defects, nor any evidence to raise question respecting the integrity of the
. core support structure.
After the refueling and inspection outage in January 1967, a hydrostatic test on the primary coolant system revealed leaks in three of the four 6-in bypass lines. The cause of leaks was attributed to intergranular. stress corrosion.
Pipe containing the defects was removed and reolaced with Type 304L stainless steel.
We conclude from our review of tt'e Litersee Event Fe: orts and the BWR Nuclear Power Exoerience that the integrity of tne reactor internal compon nts was oegraceo oy tne use of sensitized stainless steel.
The inservice inspection and testing procedures detected failures in the control rods and the 6-in.
byoass lines.
Failure of these components was attributed to intergranular stress corrosion. The cause of failure was corrected by replacing the sensitized material with material immune to attack y intergranular stress corrosion.
The inservice inspection rogram for the reactor internal components is being conducted during the current interval tc the recuirements of Section XI of the ASME Soiler and Pressure Vessel Code,1974 Edition, including Summer 1975 Addenca.
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 components is maintained during reactor operation.
I9M 012 We conclude from our review of the information submitted by the licensee that the materials in the reactor internal components are sensitized and are ocerated in an oxygen saturatec water environment, and that the incidents of stress corrosion cracking are rare because the total stress level is relatively low, not exceeding the 0.2% offset yielc strength at operating temperature.
In the unlikely event that intergranular stress corrosion. cracking should occur, cracks in the comconents will be detected by inservice inspection
- go
- ecures prior to cor:onent f ailure, We conclude that ti.1e integrity or ne reactor internal c:mponents w1i1 oe assurec by the inservice lhspection program conducted to the requirements of Section XI of the ASME Boiler and Pressure vessel Code, 1974 Edition, including Summer 1975~ Addenda, in compliance with Daragraph (g) of Section 50.55a of 10 CFR Part 50.
Further, we conclude that intergranul&r stress corrosion cracking in the reactor internal components is not a hazard to the health and safety of the public.