ML19249D586
| ML19249D586 | |
| Person / Time | |
|---|---|
| Issue date: | 08/20/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 7909250016 | |
| Download: ML19249D586 (7) | |
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g NUCLEAR REGULATORY COMMISSION
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August 20, 1979 Mr. T. M. Anderson, Manager Nuclear Safety Department PWR Systems Division Westinghouse Electric Corporation Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Anderson:
We are reviewing your Topical Report WCAP-9398, " Steam Generator Retubing and Refurbishment" submitted by your letter dated January 2,1979. We find that we require additional infomation in order for us to continue our review. details additional information which we need.
It is anticipated that additional information requests will be forthcoming as our review continues.
Your timely responses to the enclosed requests will be appreciated.
If you have any questions, the Project Manager for this review is Don Neigr,bors at 301-492-7037.
Sincerely, /
NSW A.'Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosure:
Request for additional information 790925000 1023 i61
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WCAP-9398 STEAM GENERATOR REIUSING AND REFUR3IS W.ENT REQUEST FOR ADDITIONAL INFORMATION Section 3.2 1.
Provide a detailed description of the weld preparation and configuration, welding technique, post weld heat treatment, and NDE which will be used for the installation of the two 16" manway nozzle forgings in the lower steam generator shell.
2.
Describe the technique that will be used to remove the tube-to-tutesheet welds.
Section 3.4 Provide a detailed desc*ipticn of the weld preparation.and configuration, welding technique, post weld heat treatment, and NDE which will be used for the installation of the manway access ports in the sec:ndary side of the steam generator.
Section 3.5 Provide a detailed descripticn of the weld preparation and configuration, welding technique, post weld heat treatment, and NDE which will be used for the re-installation of the upper shell assembly and steamline.
Section 4.2.1.2 1.
Is the tube expansion process mechanical or hydraulic?
2.
Discuss the potential for springback following tube expansien and possible crevice formation as a result of different elastic properties of the tube and support plate materials.
Section 4.2.1.4
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Describe the proposed. heat treatments and complete experimental results supporting the conclusion that these heat treatments can result in a sig-nificant increase in resistance to stress corrosien cracking.
Section 4.2.1.5 aha ::ntrols will be maintained to assure that tubes aill 90t be expanded Oeycn :ne tu:esneet and what wculd be the potential f:r c:rresion Of a tube ex:an:e: :eycnd the tubesheet?
Se::' - 4.2.1.6 a: I ::nei alicy aill :e se: f:r :ne ir.ternal :::a::<.n :i:e?
~. Or:.i: a seneratic of tne :i:nd:wn pi:e an cesi;n :e: ails Of the :iowc:wn iine fattenir.; te:ned an: Taterials.
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. Section 4.2.1.7 Document the baffle plate surf ace conditions, etpecially in the vicinity of the clearaneg holes and the edge of the center cut-cut.
Section 4.2.1.8 Provide a detailed descriptien of the experimental programs and results which supplied the corrosion characteristics of SA-240 Type 405 stainless steel.
Include details of the testing environments including water chemistry and temperature.
Section 4.2.1.10 Illustrate the positioning and methed of mounting the tube lane blocking devices. What materials are used for the blocking device and m unting components?
Se: tion 4.2.1.12 Frovide criteria which will be adhered to in making decisiens en the several Optiens noted in this section.
Section 4.2.1.13 Illustrate the configuration of the wrapper to shell lateral suppcrt blocks and contrast their number and design to the original lateral support blocks.
Section 5.0 Demenstrate that the methods and decontamination solutions used will not degrade or adversely affect the reactor coolant piping or com;cnents which are part of the primary system boundary. Further shcw that the decantamin-ation solution will not have deleterious latent effects in subsequent plant operation.
Section 5.3.2.7.
Section 5.3.2.7 indicates that the tube stubs will be pulled cut of the tube-sheet fr:m the secondary side.
Discuss the removal pr: cess including the dislodging cf the rolled portion of the tube frem the ucesheet and pulling of his section through the tube hole. What are the effe:ts On the tutesheet?
Es::ics 5.1 Is a:AP-5370, Revisicn 5; 'n ::nformance with ANSI N15.2-li71, "Quali y
-ss.rance Fr: gram :acuir rer.:s fcr .: lear P:wer lar s" Ir.: has i~ re:eived
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Regulatory Guide 1.50 specifically states that Secticn IX, ASME Code welding procedure qualifications,are not adequate and that the additional requirement to qualify welding procedures at the minimum preheit temperature is necessary.
Indicate your intent to comply with position c.l.b as stated in Regulatory Guide 1.50 or previde justification for an alternate positien.
4 Is WCAP-8370 in conformance with ANSI N45.2.5-1973 as required by Regulatory Guide 1.54 and has it received NRC review and approval?
5.
Regarding plugging criteria for degraded steam generatcr tubes, the tube plugging criteria must be determined on a plant specific basis and requires f;RC approval. Proposed changes in tube plugging criteria will be evaluated by the NRC in accordance with Regualtory Guide 1.121.
6.
Does WCAP-8370, Revision SA conform with the requirements of ANSI N45.2.1 as required in Regulatory Guide 1.37 and has it been approved by the NRC7 7.
Indicate the degree cf compliance with the reccamendat.icns contained in Regulatory Guide 1.44, Control of the Use of Sensiti:ed Stainless Steel (May, 1973), and 1.71, Welder Qualification For Areas of Limited Accessibility (December,1973).
Secticn 7.2.1.1 1.
What are the required tube to tube hole tolerances necessary to ensure a successful tube roll which will maintain its integrity?
2.
Clarify the discussicn en required hydrotests follcwing retubing.
It is our interpretation that a Section XI hydrotest will be required folicwing the modificatien.
3.
Expand the description of the tubesheet analysis. Describe the development of the equivalent solid plate and its properties and describe the meaning of an interaction program.
4.
Justify consideration of primary stresses only.
Is thermal shock a signif-icant concern?
5.
Describe in detail the methodology for performing the fatigue evaluation of the tubesheet.
6.
Discuss tne a; plica:ility and use of Section III, *:cnmanda:Ory Appendix A, Ar-icle A-3CCC cf the ASME Code.
Se :i:r 7.2.1.2
~ e frt: tare e:hanics ar.alysis :resente: :: :ete ' e tre 111:wable flaw 5":es is Jnl::i::' Ole. Irdi ment :f the tube s',es: Is a s:;.c : ennus
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4 Section 8.1 i
The scope of.the transient / accident evaluation in the proposed topical report is limited by the following assumptions:
the licensing regulations and guidelines in effect at the time a) of the original license are tssumed to apply, and b) enly changes in the safety analyses due te equipment changes are censidered.
As a result of these assumptions, the evaluations in the proposed topical are limited tc comparisons to the FSAR, to show tha; ac:ident/ transient analyses presented in the FSAR remain valid with tha refurbished SG design.
Our review indicates that the current Technical Spe:ifications, licensing regulations, and ECCS analyses are not necessarily limited to the regula-l license.
Our
.icns and ;;idelines in effect at the time of th,e origina:he s:c;e of t pcsiti:n is tha:
1d apply sh:uld be b catened such that plant-specific evalua-i:.s w:
licensing reculations, cuidelines and Technicai 5;e:ift:sti:ns in effect
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For pctential f:r a :lar at the tire cf steam cenera cr refurbiss ent.
Techricti 5:ecificati:n :hanges due :0 SG refur:is ent, the evaluation cf acciden s shculd refl ect u;-te-date acciden: ana'ysis m;tels and re-ce: ermine the im;act quirements; and the evaluatien of transients.shoul:
en the appr:priate reference :ycle(s), not necessarily the FSAR alone.
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Likewise, ".he evaluation of potential unreviewer. safety questions due to 55 refurbish ent should be base.1 en up-to-date regulatiens and use licensing guidelines in effect for a plan;. at the time Of a;:licatien.
- 2. The LOCA evaluatien is based on a comparisen of EC;5 performance with the Hewever, in ecst cases the refurbished and original steam generators.
FSAR ECCS analysis using the original steam genera:crs is based on a model Therefore, such FSAR analyses which the staff no lenger finds acceptable.
(or comparative evaluations) cannot be used to satisfy the requirements of Also, the ECCS analysis current at the time of steam generator 10 CFR 50.45.
refurbishment would probably have been performed assuming a significant If credit is to be taken for the number of olugged steam generator tubes.
unplugged configuration of the steam generator, a new LOCA analysis perfo For these reasons, we find that with the currently approved model is needed.
a LOCA analysis perfermed with the currently (at the time of application) a::r:ved : del must be submitted on a : Tant specific basis prict to operation The :::ical re: rt sh:uld be supple-wi-. the -t'u-ti s h e: stea genera :rs.
i-st : -ef e:: :P's eed #:r :l ant-s;e:i fi: I;:S a alysis.
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The topical' report should be supplemented with a proposed methodology for performing the plant-specific review discussed in Section B.1 which assures the applicability of the generic evaluation.
This.eth:delegy should provide guidelines for the plant-specific review and should establish criteria for determining t' hat the generic evhluation is applicable. We expect that this methodolcgy w :.1d include the following:
(a) a determination that each steam generator ; art: eter is within the expected limits identified in the te;ical re; rt (3.(c) above),
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(b) a de ermination that the generic conclusi:n f:r each transient evaluation is valid considering plant unique design and analyses, and (c) a description of requirements for submitting the plant-specific evaluation for our review.
er example, we w:uld expect that a deviation fr:m da er Ob would be identified and resolution dis-b cussed by submittal, ti:n 2.2 s
Since the refurbished units will have fewer, tubes -han the original units
, ere will be a reducticn in steam generator flew area, assuming tube diameter rema ns constant.
')iscuss how the increhsed resistance (from decreased flow area;j will alter steam venting from the :cre durine reficed and affect a:cident analysis?
Section 6.3 The topi:a1 report addresses the currently known differences in the refurbished steam generater des yn parameters but does not provide limits f:r these expected changes. Also, other steam generat:r parameters that potentially could affect the transient analysis are not addressed.
For these reasons, the proposed topical should be supplemented with the i
following:
(a) ' identify all steam generator parameters that could affect the transient analysis,
- r: vide a ;uantitative estimate of the er:ec ed P.han;e in each (b}
- ara eter f
- r the refurbished cc-diti n :?:1-ed :o the Original c:.di-ics (## a gra eter is : ex:s te-
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6-Section B.3.1 l For startup of an inacti te reactor coolant leop, confirm that all plants assume ficw in the inactive loop accelerates to its nominal full flow value instantaneously, or discuss how the analysis is affected by the lower re-sistance to ftpw in the primary side. Also, discuss the startup of an inactive leep from a configuration with the leep ste; valve initially closed.
Section 3.3.2 The evalustien cf a reduction in feedwater enthalpy states that the acci-dental opuing cf the feecwater bypass valve which diverts flow around the icw pressure feedwater heaters is an extreme ext ple cf excess heat removal by the feedwater system.
Excessive feedwater transients caused by accidental full opening of a feedwater centrc1 valve are not discussed.
Ccnfir-that the accidental c;ening of the feedwater heater bypass valve is the limiting FSA?t analysis for reduction in feed.tter enthalpy events fer all plants Or discuss cther events that are lid:i n;.
itetien 2.3.6 Fcr a stear'.ine treak, the repce: states that cne acte fer safety injection syster-actuatien is pressurizer lew pressure coincican ith low pressurizer level.
Previde an update for this mode of actuatica., censidering low pres-suri:er level signal rencval after the TMI event.
- evide a.ny additional updati"g necessary as a result of other changes im;ltmented or anticipated as a resuit cf followuc to the TMI event.
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