ML20010E464

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Forwards Response to NRC 810728 Request for Addl Info Re Fsar.Response Addresses Questions 123.2,123.3,123.4,123.7 & 123.10.Info Will Be Incorporated Into Next FSAR Revision
ML20010E464
Person / Time
Site: Wolf Creek, Callaway  Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 08/31/1981
From: Petrick N
STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To: Harold Denton
Office of Nuclear Reactor Regulation
References
SLNRC-81-078, SLNRC-81-78, NUDOCS 8109040153
Download: ML20010E464 (11)
v • d • e


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SEP 0 3 881 > 1 noNI.$rvi n I20ss0 (301)869 8010 \ "Wm7 g August 31, 1981 ,

SLNRC 81- 078 FILE: 0541 SUBJ: NRC Request for Additional Infor-mation - Materials Engineering Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos: STN 50-482, STN 50-483, and STN C0-486

References:

1. NRC (Tedesco) letter to UE (Bryan), dated July 28, 1981, same subject.
2. NRC (Tedesco) letter to KGE (Koester), dated July 28, 1981, same subject.
3. SLNRC 81-69, dated August 14, 1981, same subject.
4. SLNRC 81-74, dated August 26, 1981, same subject.

Dear Mr. Denton:

References 1 and 2 requested additional information for the SNUPPS FSAR. Reference 3 provided responses to questions 123.1, 123.5, 123.8, 123.9, and 123.11. Reference 4 provided the response to question 123.6. The enclosure to this letter provides the responses to the remaining questions, 123. 2, 123. 3, 123. 4, 123. 7, and 123.10.

This enclosure will be incorporated in the next FSAR Revision.

Very trul yours,

[\Nicholas A.RPetrick O CT V RLS/bds/10bl cc: J. K. Bryan UE G. L. Koester KGE D. T. McPhee KCPL W. A. Hansen NRC/ CAL (6 8(

T. E. Vandel NRC/l;C pg

(

8109040153 810831 PDR ADOCK 05000482 A pon

SNUPPS Q123.2 Indicate whether the individuals performing the fracture toughness tests were qualified by training and experience and whether their competency was demonstrated in accordance with a written proce-dure. If the above information cannot be provided, state why the information cannot be provided and identify why the method used for qualifying individuals is equivalent to those of Paragraph III.B.4 Appendix G, 10 CFR Part 50.

RESPONSE

The fracture toughness tests for Callaway Units 1 and 2 and Wolf Creek Unit 1 reactor coolant pressure boundary components were performed by qualified operators in accordance with written procedures.

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SNUPPS Q123.3 To demonstrate compliance with the beltline material test require-ments of Paragraph III.C.2 of Appendix G,10 CFR Part 50:

a. Provide a schematic for the reactor vessel showing all welds,
p. <tes and/or forgings in the beltline. Welds should be identified by shop control number, weld procedure qualificatial number, the heat of filler metal, and type and batch of flux.

Provide the chemical composition for these welds (particularly)

Cu, P, and S content).

b. Indicate the post-weld heat treatment used in the fabrication of the test welds.
c. Indicate the plates used to fabricate the test welds.
d. Indicate whether the test specimen for the longitudinal seams were removed from excess material and welds in the vessel shell course following completion of the longitudinal weld joint.

RESPONSE

Figure 123.3-1 identifies the location of the beltline materials and welds for the Callaway Unit I reactor vessel. Weld identification information for these welds is given in Table 123.3-1.

Information concerning the fabrication and post-weld heat treatment of the surveillance test specimen weld is identified in WCAP-9842 for Callaway Unit

1. Similar information will be provided in the surveillance program WCAPs .

for Callaway Unit 2 and Wolf Creek Unit I at a later date.

The test weldment is f abricated as a separate weld, not as an extension of a longitudinal weld seam.

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SNUPPS Q123.4 To demonstrate compliance with the fracture toughness requirements of Paragraph IV. A.1 of Appendix G,10 CFR Part 50:

a. Provide the RT for all RCPB welds which may be limiting for operation cYD[he reactor vessel.

D. Indicate whether there are any RCPB heat-affected zones which require CVN impact testing per paragraph NB-4335.2 of the 1977 ASME Code. Provide CVN impact test data for these heat-affected zones which may be limiting for operation of the reactor vessel.

c. Indicate that there are no ferritic RCPB base metals other than in vessels which require fracture toughness testing -to NB-2300 of the ASME Code. If there are ferritic RCPB base metals other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code, provide CVN impact and drop weight data for all materials which will be limiting for operation of the reactor vessel.

RESPONSE

Charpy V-notch test data for the heat-affected zone of the limiting beltline region plate is presented in WCAP-9842 for Callaway Unit 1. Similar infor-mation will be provided for the limiting materials of Callaway Unit 2 and Wolf Creek Unit 1 at a later date.

There are no other heat-affected zones which require impact testing per Paragraph NB-4335.2 of the 1977 ASME Code.

There are no ferritic base metals other than in vessels in the reactor coolant pressure boundary.

SNUPPS Q123.7 Provide full CVN impact curves for each weld and plate in the beltline region. Provide the data in tabulated and graphical form.

RESPONSE

Complete Charpy test results for each weld and plate in the Callaway Unit 1 reactor vessel beltline region are provided in Tables 123.7-1 through 123.7-3. Similar information for Callaway Unit 2 ano Wolf Creek Unit I will be provided at a later date.

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SNUPPS Q123.10 Indicate the normal operating temperature of the flywheels and provide CVN impact and drop weight test data from gach flywheel that indicates the RT of the flywheels are 100 F less than T

theirnormaloperatingNkmperatures.

RESPONSE

As stated in WCAP-8163 (Reference 1 to Section 5.4), the normal gperating temperature of the reactor coolant pump motor flywheels is 120 F. The Westinghouse of 10 F as discussed in Section specifications 5.4.1.5.2.2. The require Charpy a V-Notch maximum and RT"O[opweight gests confirm that the normal operating temperature is in excess of 100 F above the RT NDT of the flywheel material.

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