ML19220C310
| ML19220C310 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/24/1975 |
| From: | Washburn B Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904300164 | |
| Download: ML19220C310 (6) | |
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4 UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHINGTON. O.
C.
20555 March,29, 1975 Decket Nc. 50-320 A?PLICANTS:
Metropolitan Edison Company, et al.
FACILITY Three Mile Island Nuclear Station, Unit 2 (TMI-2)
SUUARY OF MEETING HELD ON MARCH 6, 1975 To discuss MATERIALS ENGINEERING 3 RANCH QUESTICNS The staff discussed, with the applicants ' representatives, our requirements for additional information neaded for adequate responses to materials engineering questions through the first-round.
A list of attendees is attached.
Significant discussion points are surrarized below.
1.
Reactor vessel materials f ractu re touchness data (0-1:
12.3 12.4 and 12.10).
The applicants' representatives outlined their program for complying with 10CFR50, Appendix G and discussed the current status of the fracture toughness informa-tion.
The applicants stated that the necessacy materi-als testing would be conducted by 3&W.
The applicants indicated that a preliminary materials availability survey had been conducted and several pieces of material are not available for testing.
The staff stated that engineering justification would be required for materials which are unavailable.
The applicants stated that they have a tabulation of available material.
The staff stated that the appli-cants should submit a schematie drawing showing where these materials are located in the vessel.
The applicants stated their intention to submit the testing program description in the next amendment, April 4, 1975, however, they do not anticipate that the materials testing results will be availaole to meet the current safety review schedule, cwdr#CS N
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. Surmary of meeting held on V, arch 6. 1975 ---
The applicants present schedule for fracture tougn-ness data ist 31W testing complete 9/1/75 Preliminary data 10/1/75 Report completed 12/30/75 The staff stated that this schedule is not compatible with the review schedule.
We require this infornation for the SER and in support of the TMI-2 Tecnnical Specifications which are scheduled for development in conjunction with the safety review.
In support of our continuing review and to permit closing this item in the SER, the applicants proposed the followings a.)
program for the fracture toughness testing to be submitted April 4,
- 1975, b.)
Using the available fracture toughness data, pressure-temperature limits, heat-up and cool-down curves will be developed on the basis of our Standard Review plan for older plants and submitted on May 30, 1975 as an interim action and c.)
i aceed with fracture toughness tests and try to submit results in time for the October 10, 1975 SER target date; if this cannot be met, the FSAR will be revised wnen the data are available.
The staff stated that this proposed approach will permit our review to proceed, however, we do not have the Charpy V-notch upper shelf toughness.
The applicant
- stated that they know of no way to obtain this until the planned testing results are available.
The staff stated that they would review the revision of the fracture toughness cata, when available, and revise the temperature-pressure limits, as appropriate.
The applicants stated that they were aware that the above interit action would result in more restrictive limitations than they would like, 0 c) \\q/
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. Summary of meeting neld on March 6,1975 ---
2 Raac or vesnel material surveillance croiram (0-1: 12 L).
The staff stated tnat S&W topical report SAW-10100 nad been approved.
The staff recognizes that samples of all f.aterials may not be avai;able; nowever, we do regttire those samples which we ballove to be the most limiting.
The staff sta*ed that we neet engineering justification for materials which sre unavailable.
The staff stated inat we need, at this stage of tne review, the program detail including criteria and basis and when the material testing is complete, we need identificati6n of samples, number of capsules, number and types of specimens, results of fracture toughness tests and chemical analyses.
The applicants stated that
- hey would submit informa-tien on the reactor vessel surveillsnee program on April 4 1975.
3.
Inservice insnection of steam zenerato.r tubes (0-1: 12. _0 ).
The applicants' respcnse to Q-1: 12.9 states that a commitment to the 1004 baseline inspection and the ex-tensive annual inservice inspectione as specified in Regulatory Guide 1.83 is not warranted for thsea steam generators.
This is based on their belisf tha t factors which have contributed to tube degradation in U-tube steam generators are not present in TM'-2 &nd personnel exposure to radiation during the inspection operation is unwarranted.
Also cited in support of this response is the lack of evidence of serious tube degradation in other units prior to, or following, the field hydro-static test.
The staff reaffirmed its cor.cerns and the need for min-imizing the probability of cteam generator tube failures as stated in Regulatory Guide 1.83.
The staff stated that they have no evidence that the steam generator tubes will not crack.
Experience statistics to date are inadequate justification for stating that the 3&W steam generators will not need inspection.
The staff stated that we need an engineering justification that inspection is not warranted and for any inservice in-spec
- ion program not providing assurance of safety
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. Summary of meeting held on Xarch 6, 1975 ---
comparable to that of the program recommended in Regulatory Guide 1.83.
The staff also stated that we will require inclusion of steam generater inc;ec-tion in the technical specifications.
The applicants stated Ine belief that secondary leak detection would be adequate sensing for cracks.
The staff stated that inspection should identify weakness prior to cracking while leak detection would provide indication subsequent to cracking: cracks and wastage are two different problems.
The staff reaffir=ed their concern that weak tubes may rupture during a iOCA or a steam line br eak.
The applicants asked the staff for additional specific comments on their respense in Amendment 22 The staff stated that the discussion pertaining to chemistry, chemistry control and chemistry effects should be quantitative and the discussion of relative stresses should also be quantitative.
The staff also commented that the stress relieving argument is not universally accepted; there is some belief that this sensitizes the material.
It is not apparent to the staff that complete steam generator stress relief results in more resistance to stress corrosion cracking.
The applicants asked the staff to document the inade-quacy in their Amendment 22 response to this question.
The staff stated that in addition to the above dis-cussion they would address a specific second-rcund question to the response in Amendment 22.
The staff stated that Regulatory Guide 1.83 is being revised and that this revision is censidering cctments made on the current version.
Tne staff stated that they would appreciate any engineering information on 3&W steam generators that would improve this guide.
The staff stated that, in consideration of the ongoing revision of Regulatory Guide 1.83 and the potential for impact of this on TMI-2, they will not further specifically address the current version of this guide in this question until the applicability of the revision is es tablished for TMI-2.
The applicants, however, should respona to the above staff concerns.
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. Summary of meeting held on N. arch 6, 1975 ----
4 Inservice instection of reactor coolant tumo flv-wneels (;-1: 12 c).
The applicants' representatives briefly outlined the testing and inspection program initiated following the CP stage ACRS letter.
Included were fracture mechanics analysis, Charpy V-notch tests, application of codes and standards and complete UT inspection following machining.
The applicants' representatives presented sketches of the pump motors and flywneels which they stated would be submitted in their expanded response to this question on April 4, 1975.
The staff stated that the information outlined at this meeting was necessary for their evaluation of the pump flywheel integrity and proposed inservice inspection.
5.
Inservica insnection cf class 1. Class 2 and class 1 systems and contonents it-1: 12,3).
The applicants submitted their proposed ESF inservice inspection program in Amendment 25, February 28, 1975, subsequent to the announcement of this meeting.
The extent of ISI for Class 2 and 3 is described in Amendment 25.
The applicants asked for our concurrence with their judgment of practicality for ISI of Class i since they interpret this to be the requirement of the propcsed revision to 1CCFR50.55(g).
The staff stated that they will review and evaluate the information submitted February 23, 1975 in Amendment 25.
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- c. W. Washburn, Project. Manager Light Water Reactors Branch 2-2 Division of Reactor Licensing 89 i;9
4 At..chment ATTENDEES TMI-2 MEETING WITH MATERIAI,3 ENGINEERING 3 RANCH MARCF 6, 1975 NAME AFFILIATION D. G.
Slear GPUSC J. M. Vann GPUSC T. M. Crimmins GPUSC E. G. Wallace GPUSC J. A. German MPR W.
S. Hazelton NRC:TR V. S. Goel NRC:TR C. D. Sellers NRC:TR
- 3. W. Washburn NRC:DL
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