ML20214G479
| ML20214G479 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/18/1972 |
| From: | Maccary R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0111, CON-WNP-111 NUDOCS 8605220188 | |
| Download: ML20214G479 (7) | |
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Roger 3. Boyd, Assistant Director for Boiling Water Rasetors Directorate of Licensing WAhMINGTON PUBLIC POWER SUPPLY 3YgTEM, HANPORD No. 2 NUCLEAR POWER PLANT, (CP), DOCKET NO. 50-397 Plant Name:
Hanford No. 2 Nuclear Power Plant Licensing iltage: CP Dc,cket Number: 50-397 Responsible Branch and Project Manager:
S. Miner Gas Cooled Rasctors Branch; Requested Completion Date: August 18, 1972 Applicant's Response Date Necessary for Completion of Next Planned Action on Project: None Description of Response: Safety Evaluation Review Status: *astomplete The information submitted by the applicant, including Amendment No has been received and evaluated by the Materials Engineering Branch
. 9, Directorate of Licensing.
that could be completed are enclosed.Our sections of the Safety Evaluation Report In order for us to complete our safety evaluation, the applicant should provide, prior to the ACRS meeting,a statement that he will comply with the requirements of Appendix G, " Protection Against Non* Ductile Fail of the recently revised ASME Code, section III, fracture toughness rules ure "
(Code case 1514), to establish pressure and temperature limitations of the reactor coolant system.during operational startup and shutdown and durin R. R. Maccary, Assistant Director for Engineering Directorate of Licensing
Enclosure:
Materials Engineering Branch CP Safety Evaluation for Eanford 2 cr.
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Form AEC.S tB (Rev 9-53 i ALCM 0240
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8605220188 720818 PDR ADOCK 05000397 A
WASHINGTON PUBLIC POWER SUPPLY SYSTEM HANFORD NO. 2 NUCLEAR POWER PLANT (CP)
DOCKET NO. 50-397 SAFETY EVALUATION MATERIALS ENGINEERING BRANCH, L REACTOR COOLANT SYSTEM Reactor Vessel Material Surveillance Program A material surveillance program is required to monitor changes in the fracture toughness properties of the reactor vessel material as a result of neutron irradiation.
The applicant has shown, in response to Question 4.5 (Amendment No. 2) and in response to Question 4.42 (Amendment No. 6), that the material surveillance program will comply with the proposed AEC E 50.55a Appendix H, " Reactor Vessel Material Surveillance Program Requirements,"
and ASTM E-185-70.
The program specification is acceptable with respect to the number of capsules, number and type of specimens, withdrawal schedule, and retention of archive material. We conclude that the proposed program will adequately nonitor neutron radiation induced changes in the fracture toughness of the reactor vessel beltline material.
. REACTOR COOLANT SYSTEM Sensitized Stainless Steel Stainless steel that has been sensitized has an increased susceptibility to stress corrosion cracking.
The applicant-has shown, in response to Question 4.7 (Amendment No. 2) and in response to Question 4.43 (Amendment No. 6), that significant sensitization of all non-stabilized austenitic stainless steel within the reactor coolant pressure boundary will be avoided through materials selection and control of welding and heat treating processes. The pre-cautions will include control of welding interpass temperatures and use of ASTM A351-CF3 material having 5 percent minimum ferrite, during fabrication and erection of piping, pumps, and valves. Stainless steel components will be joined to ferritic steel components by first buttering the ferritic steel with stainless steel weld metal, then-giving the ferritic steel a post-weld heat treatment, and finally welding the annealed stainless steel component to the buttered region of the ferritic component.
We conclude that the planning to avoid sensitization of austenitic stainless steel during the fabrication period is acceptable.
> REACTOR COOLANT SYSTEM Leakage Detection System Coolant leakage within the reactor containment may be an indication of a small through-wall flaw in the reactor coolant pressure boundary.
The leakage detection system proposed for the reactor coolant pressure boundary is described in the PSAR, paragraph 4.9, and in response to Questions 4.11 to 4.17 inclusive (Amendment No. 2).
The system will include diverse leak detection methods, will have sufficient sensi-tivity to measure small leaks, and will be provided with suitable control room alarms and readouts. The major components of the system are the containment atmosphere particulate and gaseous radioactivity monitors, and level indicators on the containment sumps. Indirect indication of leakage can be obtained from the containment pressure and temperature indica. tors. We conclude that the proposed leakage detection system will have the capability to detect small through-wall flaws in the reactor coolant pressure boundary.
. REACTOR COOLANT SYSTEM Inservice Inspection Program Selected welds, weld heat-affected zones, and other critical pressure boundary components must be inspected periodically to assure continued integrity of the reactor coolant pressure boundary during the service lifetime of the plant.
The applicant has stated, in the PSAR, paragraph 4.2.6, and in response to Questions G.1 and G.2 (Amendment No. 2), that the inservice inspection program for the reactor coolant pressure boundary will comply with Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for In-Service Inspection of Reactor Coolant Systems."
We conclude that the access provisions and planning for inservice inspection are acceptable.
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CONTAINMENT Leakage Testing Program i
Leakage testing of the reactor primary ccursincent and associated systeps is intended to provide initial and periodic verification of the leak-tight integrity of the containment, a
The applicant has stated in the PSAR, paragraph 5.2,5.1, and in response to Question 5.1 (Amendment No. 2) that the primary reactor containment 1
and its components have been designed so that periodic integrated 4
leakage rate testing can be performed at the calculated peak pressure.
i Penetrations, including personnel and equipment hatches and airlocks, and isolation valves have been designed with the capability of being individually leak tested at calculated peak pressure.
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We conclude that the design of the containment system will permit con-tainment leakage rate testing in compliance with the AEC proposed
" Reactor Containment Leakage Testing for Water Cooled Power Reactors,"
10 CFR 5 50.54(o), Appendix J, published in the Federal Register on
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August 27, 1971, and therefore is acceptable.
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. ENGINEERED SAFETY FEATURES Innervice Inspection Program The applicant has stated, in the PSAR, paragraph 6.6, and in response to Question G.3 (Amendment No. 2), that access to the Group B and C fluid systems, such as the engineered safety systems, reactor shut-down systema and cooling water systems, outside the limits of the reactor coolant pressure boundary, will be provided for inservice inspection. We conclude that the planning for an inservice inspec-tien program for the Group B and C fluid systems is adequate.