ML20054G613

Forwards Addl Info Re Inservice Insp Program,Per NRC 820415 Request.Also Forwards Basis for Design of Drywell Containment Vessel Penetrations

ML20054G613
Person / Time
Site:	Quad Cities
Issue date:	06/15/1982
From:	Rausch T COMMONWEALTH EDISON CO.
To:	Vassallo D Office of Nuclear Reactor Regulation
Shared Package
ML20054G614	List: ML20054G613 ML20054G625
References
4352N, NUDOCS 8206220136
Download: ML20054G613 (9)

Text

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!) one First Nation 11 Pl.us. Chiccgo. Hlinois C Addrtss Arply tr Post Offica Box 767 Chicago, Illinois 60690 Jun e 15, 1982 i

Mr. Domenic B. Vassallo , Chief Operating Reactors Branch #2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Quad Cities Station Units 1 and 2 Response to NRC Request for Information Concerning Inservice Inspection NRC Docket Nos. 50-254/265 Reference (a): D. B. Vassallo letter to L. De1 George dated April 15, 1982.

Dear Mr. Vassallo:

Attached, in resp]nse to your Reference (a) request, is additional information cc icerning the Quad Cities Units 1 and 2 Inservice Inspection (ISI) program.

Please direct any questions you may have concerning this matter to this o f fice.

One (1) signed original and thirty-nine (39) copies of this transmittal are provided for your use.

Very truly yours, m .ffL<

Thoma s J. Rausch Nuclear Licensing Administrator Attachment ['oV7 cc: Region III Inspector - Dresden w/a tt.

Region III Inspector - Quad Cities w/att.

Dr. D . A. Ou tlaw w/a tt.

Science Applications, Inc.

1710 Goodridge Drive McLean, Viroinia 22102 8206220136 820615 PDR ADOCK 05000254 Q PDR 4)>ZN

. t Commonw2alth Edison Comprny Responsa NRC Request for Additonal Information -

Inservice inspection' Program - Quad Cities 1 & 2.

1. Relief Requests CR-1 & CR-2.

(a) Question: Have any improvements in State-of-the-Art NDE Techniques been developed since the program submittal that would permit the examination of any of the required inaccessible welds? Please discuss.

Answer: C.E.Co. has continued to keep abreast of new developments in NDE which may assist in the examination of these welds. At this point we have not identified any new or existing methodology that would permit the com-pletion of these welds. C.E.Co. has evaluated both automated ultrasonic examination and new developments in acoustic emission technique. -The pro-blems of accessibility cannot be overcome, as of yet, with automated ultra-sonic inspection, due to the physical size of the search units being too large to traverse the areas necessary for inspection. Also, C.E.Co. has spent a substantial amount of time and effort over the last several years evaluating the possible application of acoustic emission examination to these areas. To date, the problems of severe environments for the place-ment of acoustic probes, inherent noise in the vessel, and the problems with the interpretation of results have not been solved.

It is Commonwealth Edison's opinion that the design considerations in-herent in the reactor vessel plus the other ongoing evaluations (i.e.

10CFR50 Appendix G & H requirements) provides adequate assurance of the structural integrity of the inaccessible welds and assures continued safe operation. As stated previously, C.E.Co. will continue to keep abreast of new developments which may provide us with the necessary methodology to complete these inspections.

(b) Question: Do you have an ongoing material surveillance program that conforms to 10CFR50, Appendix H7 Answer: Quad Cities Station, Units 1 and 2, does have a material sur-veillance program that conforms to 10CFR50, Appendix H. The requirement to have this program has been previously documented to the NRC. The Station Technical Specifications (operating license, DPR-29, Unit 1 Docket 50-254 and DPR-30, Unit 2 Docket 50-265) identified this require-ment. See Section 3.6/4.6 of the appropriate Technical Specifications for reference to this requirement. For further clarification and changes see the Proposed Technical Specification change submitted to the NRC, Letter Dated March 31, 1982, T. Rausch, C.E.Co. to H. R. Denton, Director, Office of Nuclear Reactor Regulations. l (c) Question: What code and addenda were used in vessel design and construction?

Answer: The reactor vessels for both Units 1 and 2 at quad Cities Station were designed and built to the A.S.M.E. Boiler and Pressure Vessel Code, Class A,1965 Edition with addenda thru summer 1965 This was previously submitted via the Plant Final Safety Analysis Report, Section 4, Amendment 13, page 4.1.2. (note: this also identifies paragraphs of later addenda used and code cases applied.).

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i (d) Question: Do you have a fracture toughness surveillance program that conforms to 10CFR50, Appendix G?

Answer: As' indicated in 1(b) above a material surveillance program conforming to 10CFR50, Appendix H, has been established. This confirms the fracture toughness of the vessel material samples. The plant Tech-nical Specifications also states that the restrictions placed on hydro-static testing, heatup and cooldown and on critical core operation are in conformance with 10CFR50 Appendix G. This information has been pre-viously submitted in the form of the plant Technical Specifications (see section 3.6/4.6 Limiting Condition for Operation Bases). Addition-ally, further changes and clarifications have been submitted to the NRC in the form of a proposed Technical Specification change dated March 31, 1982. T. Rausch, C.E.Co. to H. R. Denton, Director Office of Nuclear Reactor Regulation.

2. Relief Request CR-6.

(a) Question: Describe the leak detection systems that would serve, and their proximity to, the areas in which these pipe welds are located.

Answer: The leak detection system and the requirements and limitations on operation are explained in Section 3.6/4/6 of the Plant Technical Specifications. This system consists of the drywell (Primary Containment) sumps which are monitored for identified leakage (Drywell Equipment Drain Sumps) and unidentified leakage (Drywell Floor Drain Sumps). These sumps are located at Elevation 579'10" which is the basement level of the dry-well. The following list of penetrations gives the penetration elevations:

CRD Return -0308-4" - Elev. 618' RHR - 1011-4" - Elev. 604' 1012-ASB-16 - Elev. 590' 1025-20" - Elev. 605'

, Rx Wtr Cleanup-1202-6" - Elev. 625' l Core Spray-1403-10" - Elev. 642' i

1404-10" - Elev. 642' l HPCI-2305-10" -

Elev. 590' Main Steam-3001A,B,C,D-20" - Elev. 594' Feedwater-3204A&B-18" - Elev. 598' (b) Question: Discuss the capability of the bellos sleeve to withstand the full dynamic effects of a longitudinal or circumferential break of the enclosed process pipe, including jet impingement, pipe whip impact, and environmental effects. if previously discussed in submittals to the NRC, document by references.

Answer: The design consideration of the penetrations were established using General Electric Specification Number 22A2505, Rev. O. titled " Basis for Design of Drywell Containment Vessel Penetrations". (at tached) . As further information see the attached letter from C.E. Company Station Nuclear Engineering Department.

1 (c) Question: Do any of the 15 pipe welds involve the welding of dissimilar '

metals? )

1 Answer: In all cases the enclosed wela is made between similar metals.

3 Relief Request CR-9 (a) Question: Has ultrasonic wall thickness measurement been considered as an alternative examination?

Answer: This alternative has been considered by the station, however we feel that this option is impractical for the following reasons:

(1) 'The pump pressure retaining boundary material is ASTM A351, Grade CF8M.

The inherently coarse grain structure of this cast stainless steel material results in high ultrasonic attenuation and generation of extraneous ultrasonic signals often rendering thickness measurements indeterminate.

(2) Since the reactor recirculation pumps are cast and machined assemblies, there are large areas of the pump bowl where the inner and outer sur-faces are not parallel. This geometry precludes meaningful ultrasonic thickness measurements.

(3) Personnel radiation exposure in the general area of the pump casing have been in the range of 1500 to 3000 MR/ hour. To expend the time and accumu-lated mandose necesscry to delnsulate, prepare (if necessary), inspect

. id reinsulate in order to perform thickness checks does not balance ejainst any possible gains in safety margins.

The foregoing reasons establish what we feel is an adequate justification for not implementing a wall thickness program. The program we propose (i.e. inspection of the internal surfaces to the extent practical, when disassembled for maintenance) provides assurance of the safety of the recirculation pump casing.

(b) Question: What are the manufacturers recommendations regarding the dis-l assembly of the pumps for regular maintenance?

i Answer: The pump manufacturer's (Byron Jackson Pump Division, Borg-Warner Corporation) experience does not indicate a need to regularly disassemble these pumps for maintenance or inspection. In their opinion, it is likely

} that the station would operated for ten years without at least one pump

being disassembled in each unit for maintenance.

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(c) Question: Based on Industry experience with the pumps, what is the likelihood that the Quad Cities plant will operate ten years without at least one pump being disassembled in each unit for maintenance?

Answer: To gather the information necessary to make this Judgement, the following question was asked of NOMIS (Nuclear Operations and Main-tenance information Service):

(a) Have you ever disassembled your Reactor Recirculating or Reactor Coolant Pumps?

(b) Was disassembly for maintenance or ISI work (excluding mechanical seal replacement)?

(c) Have you ever asked for rellef from the NRC concerning disassembly of your Class 1 pumps solely for the purpose of performing ISI work?

The question was responded to by twenty-five stations with the following general responses:

(1) Fifteen plant sites have not disassembled a recirc or coolant pump as of yet.

(2) One plant disassembled a recirc pump during preservice inspection.

(3) One plant disassembled coolant pumps during preop testing, for main-tenance.

(4) Four plants have disassembled one pump for maintenance purposes (not including Quad Cities).

(5) Three plants have disassembled pumps for maintenance and ISI.

(6) One plant has disassembled pumps solely for ISI purposes.

(7) Quad Cities Station has disassembled one pump for maintenance. This pump (Unit 2-2A-202) was disassembled to replace the pump bowl to cover gasket. While disassembled, the pump was visually inspected to the extent practical and no degradation of the pressure boundary surfaces was noted. This disassembly and inspection was completed in February 1978 which was approximately five years into the first inspection interval.

In our judgement, the answers provided by the industry are still incon-clusive as to the likelihood of any plant requiring maintenance on a recirc pump, which would allow visual ISI on the internal pressure boundary sur-faces. However, with the maintenance type approach, a realistic gain is achieved by the industry in that the tremendous expenditure of time and personnel radiation exposure is utilized on pumps requiring disassembly because of problems. It must be stated that even the least complicated pump requires detailed and complex dimensional checks which could affect pump operability. This detailed check is diluted when large dose rates are evident because of the time it takes to complete the readings and in-spection. Commonwealth Edison Company feels it is in the best interests

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of safety and reliable operation of the pumps to perform the pressure boundary inspections when the pumps are disassembled for maintenance.

This stance is enhanced by the necessity to maintain personnel dose rates as low as reasonably achievable.

4. Relief Request CR-10 (a) Question: Has ultrasonic wall thickness measurement been considered as an alternative examination?

Answer: This alternative has been considered by Quad Cities Station however, we feel that this option is impractical for the following reasons:

(1) The inherently coarse grain structure of the cast stainless steel valve materials result in high ultrasonic attenuation and generation of extraneous ultrasonic signals often rendering thickness measure-ments indeterminate.

(2) The point to point thickness variations in casting thickness result in difficulty in repeatability of the thickness measurements.

(3) Historically, wastage defects in castings tend to be localized. As there are large areas of non parallel surfaces where ultrasonic thick-ness measurements cannot be performed, the probability of detecting wastage defects is small.

(4) Radiation levels in the valve areas is generally high. The man-hours and radiation exposure expended to perform the examination is not commensurate with the increase in safety margins.

The foregoing reasons establish sufficient justifications for not imple-menting a wall thickness program. Quad Cities Station has, in the first interval, already examined a substantial number of valves -which were dis-assembled for maintenance purposes. The examinations have not identified any indication of pressure boundary degradation. This,pl6s the commitment to inspect all classified valves when disassembled, constitutes an adequate sampling program to meet the intent of Section XI . This program of examina-tion provides assurance of the integrity, and thereby the safety, of the valves pressure boundaries.

(b) Question: Based on industry experience, what is the anticipated mainte--

nance frequency for the valves?

Answer: Based on experience at Quad Cities Station,'It is anticipated that most groups of valves will have at least one valve disassembled for maintenance each ten years. In Quad Cities Station first nine years of commercial operation, we have had at least one valve opened in seven of nine groups in Unit I and in eight of nine groups in Unit 2.

Quad Cities feels that his experience plus the commitment in our relief request to examine each valve disassembled, rather than one of each group, constitutes an adequate sampling program to identify developing problems with valve pressure boundaries.

5 Relief Request CR-Il (a) Question: Does the design of the heat exchanger permit an internal inspec-tion of these welds?

Answer: Access to these internal surfaces is only achieved when a complete disassembly of the heat exchanger is accomplished. This type of disassembly would only be accomplished when major maintenance is required (i.e. tube bundle replacement).

(b) Question: If so, has an alternative internal examination been considered if and when a heat exchanger is opened for maintenance and/or repair?

Answer: This type of internal examination has been considered, however, the i n te r fe rence inherent in each joint would preclude performing volumetric inspection. Possibly a remote visual examination could be performed, to the extent practical, dependent upon internal radiation levels.

(c) Question: What is the anticipated frequency of maintenance for these heat exchangers?

Answer: The heat exchangers have not been completely disassembled at the site and it is not anticipated that they will over the plant lifetime. We have performed maintenance on the heat exchangers (some tubes have been plugged, drain beilows replaced, etc.) but this type of maintenance does not provide access to the internal weld surfaces in question.

It should be noted that the reinforcement saddles were fabricated with drilled vent holes. These holes provide a means to determine if the ob-structed pressure retaining weld has developed any through wall defect.

This vent hole will be observed during the required Section X1,1WC-5000 pressure tests. C.E.Co. feels that this inspection program provides for adequate margins of safety in the continued use of the heat exchangers.

6. Exemption No. 5 A revision to Quad Cities Station 151 program is being generated to delete the reference to Exemption Number 5 No relief requests will be submitted, even though there are obstructions to certain welds. The station feels confident that the required percentage of welds to satisfy code requirements can be accom-m plished without examining welds which are obstructed, therefore, no specific relief is being requested.

7 Program Interval Commonwealth Edison Company has previously submitted a letter identifying our 40 month period dates. This letter is dated July 12, 1976, G. A. Abrell, Nuclear Licensing Adminstrator, C.E.Co. to Karl R. Goller, Asst. Director, Division of Operating Reactors, USNRC.

Additionally, our commercial service dates for Unit 1 & 2 are given in NUREG 0200 (Grey Book) .

8. Relief Request CR-13 (a) Question: Please demonstrate that the acoustic velocity and attenuation of the calibration blocks that lack documentation fall within the range of the straight-beam longitudinal wave velocity and attenuation of the components to be examined.

Answer: Quad Cities station is rewriting Relief Request CR-13 to include the information on our commitnent to verify the acoustic compatibility of the vessel calibration block to the reactor vessels. CR-13 is also being rewritten to delete from the scope of the relief request the reference to more than one standard. When this Relief Request was originally written some piping standards were erroneously included in the scope. Quad Cities piping standards are in conformance with the requirements of Article 5 of Section V of the ASME Code, therefore, no relief is required for those standards.

(final)

NER AL h ELECTRIC

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NO.OP ACORESO33 l c o '

NUCLEAR ENERGY DIVISION 1 Naymark, S-353 2 Roof, RR-320 TRA M TTAL 1 Omer, JE-PLM 19 Alexander, RE-366 (QC)

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i OYSTER CREEK; 12 Pashos, TJ-359 1 Tech Pubs-509 DRESDEN 2-3; QUAD CITIES 1-2; 2 Kosut, BS-632 1 Teissier, RP-367 l l

1 Vail, DB-353 1 Ascherl, RJ-367 PROJcCT MILLSTONE; MONTICELLO 1 Fiock, WL-377 1 roarrando, NL-377 -

1 GRU-742 8+1R Alexander, RE-366 (Mill)

REQUISITION 14 Alexander, RE-366 (OC) 1 Tech Pubs-509 l Dote February 18. 1969 1 Scott, WM-353 2 Wood, LE-364 1 Bloom, TE-377 1 Hazen, RA-377 1 Tech Pitbs-509 1 Spencer, R-592 l TO: DISTRIBUTION Wolf, LW-377 19 Alexander, RE-366 (D2/3) 2

'_ 1 Lattin, NF-377 6+2R Alexander, RE 366 (Mont) ,

F ROM: D. B. VAIL i

' 1 Fogelquist, JD-367 1 Tech Pubs-509 1 Tech Pubs-509 1 Vassar, RK-350

SUBJECT:

SPECIFICATION Skarpelos, JM-359 I 1 Teissier, RP-367 1

" BASIS FOR DESIGN OF DRYWELL 1 Devine, WD-663 0 Huggins, RA-353 CONTAINMENT VESSEL PENETRATIONS" 1 Ascherl, RJ-367 TMCY AmE TOSE ATTACHED USED POR THESE ARE PLE ASE NOTE PLEASE R E QU IS IT IONS DIS T a i SU TION PR E LIMIN AR V R E WislON COMMENT P RIN T S E S T IM A TIN G UNCHECNED NOLOS APPROVE PHOTOSTATS ORDERING M AT* L CHECRED DISTRIBUTE X BPECf PIC A TIONf SECURING QUOT. X PIN AL TR ACINGS X C ON S T R UC TION .

R E PR ODUC f 9LE IN FORM A TION PROCUREMENT DR AWsN G NuM BERS, TITLES & C OMME N T S Specification 22A2505 Rev. 0 " Basis for Design of Drywell Containment Vessel Penetrations" i

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