ML19263D966
| ML19263D966 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/26/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 7904170120 | |
| Download: ML19263D966 (6) | |
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UNtTto STATES
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NUCLEAR REGULATORY COMMisslON
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WASHINGTON, D. C. 20555 t ~ A:qq,'%,,,.*f MAR 2 6 G79 Docket tios.:
50-329 50-330 Consumers Power Company ATTN:
Mr. S. H. Howell Vice President 212 West Michigan Avenue Jackson, Michigan 49201 Gentlemen:
SUBJECT:
CORRECTION TO LETTER OF MARCH 22, 1979 The enclosed requests for additional information on the FSAR for Midland Plant, Units 1 & 2, were inadvertently omitted in my letter of March 22, 1979 and were transmitted by telecopier to Mr. Jim Zabritski of your organization on March 23, 1979.
Please add these enclosed requests to those foraarded previously,
- incerely, 1
kl%Y JNCQ Steven A. Varga, Ch e6 Light Water Reactor Branch tio.4 Division of Project Management
Enclosures:
Requests 121.23 theough 121.32 cc:
See next page 7904170120 A
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Consumers Power Ccmpany ccs:
Michael I. Miller, Esq.
Isham, Lincoln & Beale Suite 4200 One First National Plaza Chicago, Illinois 60670 Judd L. Bacon, Esq.
Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Mr. Paul A. Perry Sec reta ry Consumers Pcwer Company 212 W. Michigan Avenue Jackson, Michigan 49201 Myron M. Cherry, Esq.
One IBM Plaza Chicago, Illinois 60611 Mary Sinclair 5711 Summerset Drive Midland, Michigan 48640 Frank J. Kelley, Esq.
Attorney General State of Michigan Environmental Protection Division 720 law Building Lansing, Michigan 48913 Mr. Wendell Marshall Route 10 Midland, Michigan 48640 Grant J. Merritt, Esq.
Thompson, Nielsen, Klaverkamp & James 4444 IDS Center 80 South Eighth Street Minneapolis, Minnesota 55402
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w 121-1 121.0 MATERIALS EilGIt1EERItG BRAtiCH - MATERIALS IrlTEGRITY SECTI0t1 121.23 The preservice inspection program for the Midland Plant, for (5.2.4)
ASME Code Class 1, 2, and 3 components (letter, S. H. Howell to RSP R. S. Boyd, CPCo Serial 5930, October 5, 1978, submitted in respor.se to tiRC Questions 121.1,121.3 and 121.14) is ut adequate. The proposed preservice inspection prograr, aces not reference the edition and addenda of Section XI of the ASME Code required by 10 CFR Part 50, Section 50.55a, nor the edition and-addenda of the ASME Code that will be required by the proposed change to this regulation (FEDERAL REGISTER, Vol. 44, flo.13, January 18, 1979, pp. 3719-3721).
It is our position that you submit a preservice inspection program for the Midland Plant that complies with 10 CFR Part 50, Section 50.55a.
121.24 The proposed steam generator inspection program contained in the (5.4.2)
Midland Plant FSAR and Technical Specifications, is not acceptable.
(16.0)
RSP Response to Question 121.3 indicates that a preservice eddy current and ultrasonic examination and inspection of the steam generators has been conducted. Provide a description of the preservice inspection and a summary of the inspection results (FSAR Section 5.4.2 and Technical Specification Section 16.3/
4.4.5).
It is our position that Section 16.3/4.4.5 (applicability and bases) of the Midland Plant Technical Specification be revised to be consistent with the corresponding section of t;UREG-0103,
" Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors."
121.25 Midland Plant FSAR Table 5.2.3 indicates that SA-533 Grade 8 (5.2)
Class 1 plate material is used in the fabrication of flSS-12 and flSS-13 reactor vessels.
Your response to flRC Question 121.10 indicated no use of this material. Clarify this discrepancy.
121.26 Table 5.2-1 of the Midland Plant FSAR indicates that components (5.2) of the reactor coolant pressure boundary were ordered and (5.3) constructed to editions and addenda of the ASME Code that were effective prior to the issuance of 10 CFR Part 50, Appendix G.
Section 5.3.1.5 of the FSAR discusses the difference in fracture toughness requirements between the ASME Code and 10 CFR 50, Appendix G.
Table 5.3-2 of the FSAR lists fracture toughness test results for the materials of the reactor beltline region and lists the estimated fracture toughness for materials in other areas of the reactor vessel and other components in the reactor coolant pressure boundary. Babcock and Wilcox Topical
121-2 Report BAW-10046A, " Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR Part 50, Appendix G," was used to provide the estimated fracture toughness values in Table 5.3-2.
Sections III and IX of the ASME Code require mechanical testin<j of materials to be used throughout the reactor coolant pressure boundary, not only beltline region materials. To demonstrate compliance with Appendix G to 10 CFR Part 50 and to demonstrate applicability of the estimations of BAW-10046A, supply tha results of the ASME Code required tests (i.e., test required, ASME Code paragraph, yield stress, ultimate tensile stress, impact energy, lateral expansion, test temperatures) for all of the ferritic materials used in the reactor coolant pressure boundary.
Identify any of the reactor coolant pressure boundary material test results that were obtained prior to developing standard documentation to demonstrate personnel competency in materials testing (Paragraph II.B.4, Appendix G to 10 CFR Part 50, item of non-compliance identified in CPCo response to ttRC Question 121.17).
121.27 Paragraph IV.A.4 of Appendix G,10 CFR Part 50, requires that (5.3) all bolting over one inch nominal diameter meet a minimum of 25 mils lateral expansion and 45 foot-pounds as determined by Charpy V-notch tests.
Section 5.3.1.7 of the Midland Plant FSAR presents tensile strength and Charpy V-notch energy data for the reactor vessel fasteners only.
Confinn that the reactor vessel fasteners are the only bolting over one inch ncminal diameter, or supply the required test results (and acceptance standards used if different from Appendix G) for any other bolting material in this size classification. As specified by Appendix G to 10 CFR Part 50, bolting includes bolts, nuts and washers.
Identify any of the bolting material test results that were obtained prior to developing standard documentation to demonstrate personnel competency in materials testing (Paragraph II.B.4, Appendix G of 10 CFR Part 50, item of non-compliance identified in CPCo response to ftRC Question 121.17).
121.28 Babcock and Wilcox Topical Report BAW-10056A, " Radiation
- (5.3)
Embrittlement Sensitivity of Reactor Pressure Vessel Steels,"
dated August 1973, is referenced in Section 5.3.1 of the Midland Plant FSAR. This report presents background information and materials test results that were used to fomulate radiation damage curves.
In July 1975, the ftRC issued Regulatory
121-3 Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," (Revision 1 issued April 1977) which presents radiation damage curves acceptable to the fiRC staff.
In response to f4RC Questions 121.5, 121.12, 121.18 and 121.21, CPCo has ccanitted to fully implement the recomendations of this regulatory guide for the Midland Plant.
Consequently, Babcock and Wilcox Topical Report BAW-10056A is not applicable to the Midland Plant licensing review and reference to this topical report in the FSAR should be deleted.
121.29 To demonstrate compliance with Appendix H to 10 CFR Part 50, (5.3) include in the Midland Plant FSAR and Technical Specifications a (16.0) table that provides the following infonnation for each surveil-lance specimen capsule:
(1) The actual surveillance materials in each capsule.
(2) The test specimen type (s) made frcm each material.
121.30 Revise Table 5.3-7 of the Midland Plant FSAR to show the follow-(5.3) ing for each surveillance specimen capsule:
(16.0)
(1) Proposed loading schedule of capsules into the reactor vessels.
(2)
Indicate the specific surveillance capsules that will be placed in the locations identified in Figure 5.3-6.
(3) Proposed time of capsule withdrawal (calendar years and effective full power years).
Incorporate this table into the Technical Specifications for the Midland Plant (Table 4.4-5).
121.31 Babcock and Wilcox Topical Report BAW-10100A, " Reactor Vessel (5.3)
Material Surveillance Program, Compliance with 10 CFR 50, Appendix H, for Oconee Class Reactors," dated February 1975, is referenced in Section 5.3.1 of the Midland Plant FSAR. This report presents discussions on surveillance specimen capsules, surveillance specimen holder tubes, neutron flux lead factor, radiation damage, holder tube mounting locations, surveillance specimen types and number.
Due to operating problems experienced by the surveillance specimen capsules and holder tubes, Babcock and Wilcox has redesigned the holder tubes and changed the mounting locations resulting in different neutron flux lead factors. The capsule
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121-4 itself has bean redesigned to hold different surveillance specimen types and quantities. Also, as discussed in t;RC Question 121.28, the radiation danage curves as presented in this topical report, and in BAW-10056A, are no longer used in the Midland Plant FSAR.
Consequently, Babcock and '.filcox Topical Report BAW-10100A is not applicable to the Midland Plant licensing review and reference to this topical report in the FSAR should be deleted.
121.32 Figure 4-1, " Fast fleutron Fluence (E > 1 MeV) as a Function of (16.0)
Full Power Service Life," Figure 4-2, " Eft ~ect of Fluence and Copper on Shift of RT. T for Reactor Vessel Steels Exposed to 550 F Temperature," ad' Table 4-1, " Reactor Vessel Toughness,"
of the Midland Plant Technical Specifications have been lef t blank. Supply this information.
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