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s BAR.: 21979 Docket Nos. 50-329/330 MS 12-12 IENDR WISIPUR:
- 3. A. Varge. Chief Light Water Reactors treach No. 4 Divistas of Project Manegament FRON:
- 5. 3. Paulitti. Chief Raterials Esgineering Branch e
SUBJECT:
CONSUMERS POWUt COWANY (CPCo). MIRAND PLAE. UNITS 1 AND 2 Plant Name: Midland Plant Suppliers: Babcock and Wilcan; Bechtel Licenstag Stage: Ot.
Docket m sers: 50-329/330 u
ansponsible Branch and Project Manager: Udt 4; D. Hood noviamer:
N-L. Boyle nascriptim of Task: 0-2 saeplement
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Review Status: Information Required In a ammorsedum from 5. 5. Paulicki to 5. A. Yarge, dated Jaamary 17, 1979, the Materials Integrity Section. Materials Engineering Branch. Divistem of Systems safety. identified four revier armes that required additiemal toforestion from CPCs before the safety evelatius for the Midland Plant.
N could be coupleted. NTIB ressived Revision 18 to the Nidland Plant F5HR an March 2.1979 and un have avslueted the.tafbruettes contaiend is this
.Y revista. Me have reached a camelusten that tuo of the areas iden
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are predicties of radiaties dange to RV motorials and efenets of atypteet
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sold startet,os preneuretemperature limits.
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He have also concloded that the following too areas require additiemal inferention before me muy cumplete our review of the materials istogrity af the Ridland Plast.
1.
Non-Cameliance with Appendices 8 and N.10 CHt 50 Certais areas of nmH:ompliance wtth these regulations have been l
identified by MTIB and by CPCs (response to Question 121.17). The response to this questics, and the pertinent FSAR Sections, have preved to contata insufficient informaties
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....s We reoufre CPCo to prwide the infonetton identified in the attached questions so that we can complete our review of this ite.
2.
Preservice and Insenice inspectios <%.a.10 CR 50.55a CFCo semitted proposed preservice and faservice faspectics programs.
CFCo letter dated Octsher 5.1978. These programs de not reference the edition and addsada of Section XI of the ASE Code required by 10 CFR 50.55s. ner me esttien and addends that will be required ty the proposed champs to this regulatten (FDGAL RESISTD. January 18, 1975).
i Therefore, fa order to esuplete our rwiew of the preservice program.
we require CPCs to submit a rwised %.. ice inspection progrun tMt couplies wf th the requirements of 10 CFR 50.55a. Stace the inservice faspection program is required to comply with the edition and addenda of Sectfen XI of the ASME Code that is in effect six months prior to the date of commercial operation, we do not require the details of the inservice inspection program to complete our SS review.
Upon receipt of adegasta responses to the attached questions, so will preparv eur SG input and seuf t it to DPM.
CWSY s, s. PevkW
- 5. 5. Pawlickt. Chfef Materials Engineering Branch Division of systems safety Offfce of itselaar Caector llegulation 1
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121-1 121.0 MATERIALS DIGINEERING BRANCH - MATERIALS INTEGRTTY SECTION 121.23 The preservice inspection program for the Nidland 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 response ts NRC Questions 121.1.121.3 and 121.14) is not adequate. The proposed preservice inspection program does 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. No.13 January 18,1979, pp. 3719-3721).
It is our position that CPCo submit a preservice inspection 1
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)
'5P Response to NRC Question 121.3 indicates that a preservice eddy current and ultrasonic examination and inspection of the steam generators nas 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 (applicabillty and bases) of the Midland Plant Technical Specification be revised to be consistant with the corresponding section of NUREG-0103
" Standard Technical Specifications for Sabcock and Wilcox Pressurized Water Reactors."
121.25 Midland Plant FSAR Table 5.2.3 indicates that SA-533 Grade B (5.2)
Class 1 plate material is used in the fabrication of NSS-12 and NSS-13 reac:or vessels. CPCo response to NRC Question 121.10 indicated as use of this material. clarify this discrepancy.
121.26 Table 5.2-1 of the Midland Plant FSAR fndicates 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 riquirements between the ASME Code and 10 CFR 50 Appendix G.
Table 5.3-2 of the FSAR If sts fracture toughness test results for the materials of the reactor beltline region and lists tae estimated fracture toughness for materials in other areas of the reactor vessel and other components in the reactor coo' ant pressure boundary. Babcock and Wilcox Topf(.a1
e 121-2 Report 8AW-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.
r Sections III ad IX of the ASME de riquire mechanical testing of materials to be used throughai the reactor coolant pressure boundary, not only beltline regiw materials. To demonstrate compliance with Appendix G to 10 CFR Part 50 and to demonstrate applicability of the estimations of 8AW-10046A supply the results of the ASME Code required tests (i.e., test required.
ASIE Code para < aph, yield stress, ultimate tensile stress, impact energy, lateral expansion, test temperatures) for all of the ferHtic meterials 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 personeel competency in materials testing (Paragraph II.B.4, Appendix G to 10 CFR Part 50, item of non-compliance identified in CPCo response to NRC Question 121.17).
121.27 Paragraph IV.A.4 of Apperdix G,10 CFR Part 50, requires that (5.3) all bolting over one inch nominal diameter meet a minimus of 25 mils lateral expansion and 45 foot-pounds as determined by Charpf 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.
Confirm that the reactor vessel fasteners are the only bolting over one inch nominal diameter, or supply the required test results (and acceptance standards used if different from Appendix G) for any other bolting mateMal 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 staMard 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 stC Question 121.17).
121.28 Babcock and Wilcox Topical Report SAW-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 fonsulate radiation damage curves. In July 1975, the NRC issued Regulatory
121-3 Guide 1.99, " Effects of Residual Elements on Predicted Radiation f
Danspe to Reactor Vessel Materials " (Revision 1 issued April 19770 which presents radiation damage curves acceptable to the NRC staff.
In response to NRC Qwstions 121.5,121.12.121.18 and 121.21, l
CPto has coussitted to fully implement the reconmendations of this regulatory guide for the Midland Plant.
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Consequently. Babcock and Wficox 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 tae following infonnation for each survell-lane specimen capsule:
(1(2) The actual surveillance materials in each capsule.
) The test specimen type {s) made from each noterial, t
121.30 Revise Table 5.3-7 of the Midland Plant FSAR to show the follow-l (5.3) ing for each surveillance s:ecimen capsule:
(16.0)
(1) Proposed loading schedale of capsules into the reactor vessels.
(2) Indicate the specific surveillance capsules that will be placed in the locations identified in Figure 3.3-6.
(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).
i 121.31 Sabcock and Wilcox Tootcal Report BAW-10100A,
- Reactor Vessel l
(5.3)
Matarial 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 tute mounting locations, surveillance specimen types and numeer.
Due to operating problems experienced by the surveillance specimen capsules and holder tubes. Babcock and Wilcos has redesigned the nolder tubes and changed the mounting locations resulting in different neutron flux lead facte s.
The capsule
o 121-4 itself has been redesigned to hold different surveillance l
specimen types and quantities. Also, as discussed in NRC Question 121.28, the radiation damage curves as presented in I
this topical report, and in BAW-10056A, are no longe' used in the Midland Plant FSAR.
Consequently, Babcock and Wilcox Topical Report BAW-10100A is not appitcable to the Midland Plant licensing review and reference to this topical report in the FSAR should be deleted.
Sufficient information has been provided in the FSAA and other Babcock and Wilcox topical reports, and with the CPCo censit:nent to fully implement the reconnendations of Regulatory Guide 1.99, we require no additional infonnation in this area.
121.32 Figure 4-1, " Fast Neutron Fluence (E > 1 MeV) as a Function of (16.0)
Full Power Service Ltfe," Figure 4-2, "Effect of Fluence and i
Copper on Shift of RT for Reactor Vessel Steels Exposed to 550 F Temperature," agTTable 4-1, " Reactor Vessel Toughness,"
of the Midland Plant Technical Specificaticns have been left blank. Supply this infonnation.
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