ML19206B398
| ML19206B398 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/17/1974 |
| From: | Washburn B US Atomic Energy Commission (AEC) |
| To: | Kniel K US Atomic Energy Commission (AEC) |
| References | |
| NUDOCS 7905090439 | |
| Download: ML19206B398 (5) | |
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Reacter Vessel Internals
?rc' ride a list of any cold-worked austenitic stainless steels, precipitatica hardening stainless steels or hardenable nartensitic stainless steels having
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yield strengths greater than 90,000 psi used for cenpanents of the reactor vessel internals. If any such steels are usad, provide assurance that they will be conpatible with the reactor coc_ ant.
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.t. J V $4.4 J Ceneral 'daterials Considern-ion.s In Section 3.2..'.3 of the FS;_R the applicant states that the reactor coolant insulation is all retal reflectite insulation. Describe the requirements for non=etallic insulation for stainless steel cenponents important to safety, particularly with respect to chlcride, fluoride, and silicate centant. Indica :e the degree of conf ormance
-h Regulateri Guide 1.36, "'Ionmetallic Thernal Insulation f or Austenitic Stainless Steel," dated February 23, 1973. Include detailed infor ation en the nature of the control (s) of installation. Of specific interest in this respect are the types of cements that are field ci:ced for appliention.
Fracture Toughness Oceratinz Li=itations During Startua and Cooldown In Section 5.2.4.3 of the FSAR the applicant states that Appendia G to Section III was used as a guide in establishing operating pressure-temperature li=itations of the reactor coolant systes and that deviations frc= Ac. c. endi:c G are described in report 3A'J-10046.
Provide us a coav of e,
the repcrt, B24-10046, or the information contained in the report, for our review and evaluation.
In addition, the temperature-pressure limitations for core operation cust confor-to the requirements of s,,ma..
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. Reactor Vessel Material Surveillance Pro::an Provide the entent of confor ance of the reactor vessel surveillance program to ASIN-E-133-73 and Appendix H,10 CFR 30 and provida j ustification for any daviaticas.
Pu=o Flvwheel Integrity Provide sufficient inf or:stion about the pri=ary coolant pump-actor fly-wheels to indicate the degree of flywheel integrity cc= parable to the progras recc= mended in Regulatory Guide 1.14, " Reactor Coolant Pu=p Flywheel Integrity," October 27,. 1971. The inservice inspection program for flywheels presented in the Technical Specifications, Specification 4.2.3, is not satisfactory. The inspection progras should be cc= parable to the requirenents of paragraph C.4 of Regulatory Guide 1.14 and should be included in Table 4.2-1 of the Technical Specifications.
Inservice Inscection Pro 2 ram of ASMI Ccde Class 2 & 3 Ccrocnents Provide sufficient inf or=ation about your proposed inservice inspection progras to indicate that the progrma will provide a degree of assurance of syste= integrity comparable to the progran reco== ended in Regulatory Guide 1.51, " Inservice Inspection of ASMI Ccde Class 2 and 3 Nuclear Pcuer Plant Ccaponents," May 1973. Tha progra= for C?as; 2 cc=penents should include a table in the Technical Specifications using the for:at
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. TECEiICAL SPECIFICATIONS Liritinz Canditions f or Coeratica 4
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apectri:stian 3.1.g. 1, i f +.cycra te s ts,If re:ers to speci: cation
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- or pressure li itations when there are fuel assachlies in the vessal.
Specifications 2.2 and 3.1.2.1 should also proviue ta=perature linitations calculated in accordance with the require =ents of paragraph G-2410 of Appendi G to ASME Boiler and Pressure 7essel Ccde,Section III.
Provide the basis for the heatup ana cooldown li=it curves, Figs. 3.1.2-1 and 2.
This should include specific fracture toughness data used to ting initial RT and the a= cunt of radiation induced determine the 14 4 tenperature shift. The chemical analysis of the vessel belt 11ae materials, particularly those elements such as copper, phosphorus, and vanadius which affect fracture toughness properties, should be provided.
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