ML20111C292

Forwards marked-up Pages to Draft SER Needing Correction,In Response to Request for Supplemental Info.Addl Corrections Will Be Submitted When Available.Changes Should Appear in SER Scheduled for 850601

ML20111C292
Person / Time
Site:	Vogtle
Issue date:	03/11/1985
From:	Bailey J SOUTHERN COMPANY SERVICES, INC.
To:	Adensam E Office of Nuclear Reactor Regulation
References
GN-549, NUDOCS 8503150339
Download: ML20111C292 (11)

Text

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Southem Company Servces, Inc.

Post Otoce Box 2625

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• Birmingham, Alabama 35202 Telephone 205 870-6011 A

Vogtle Project March 11, 1985 Director of Nuclear Reactor Regulation File: X7BC35 Attention: Ms. Elinor G. Adensam, Chief Log: GN-549 Licensing Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 REQUEST FOR SUPPLEMENTAL INFORMATION DSER COMMENTS

Dear Mr. Denton:

Enclosed for your information are two (2)' sets of those pages of the VEGP Draft SER which we feel should be corrected, as indicated by the written in changes, in the SER scheduled for June 1, 1985. Additional corrections will be submitted as they become available.

If your staff requires any clarification on our corrections, please do not hesitate to contact me.

Sincerely, J. A. Bailey Project Licensing Manager JAB /sp xc: D. O. Foster R. A. Thomas G. F. Trowbridge, Esquire J. E. Joiner, Esquire C. A. Stangler  ;

L. Fowler l M. A. Miller i L. T. Gucwa G. Bockhold, Jr.

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t in the GDC 27 and 35. Specific coolability requirements for the loss-of-coolant accidents are given in 10 CFR 50.46 (" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors").

To meet the above-stated objectives of the fuel system review, the following specific areas are critically examined: (a) design bases, (b) description and design drawings, (c) design evaluation, and (d) testing, inspection, and surveillance plans. In assessing the adequacy of the design, several items involving operating experience, prototype testing, and analytical predictions are weighed in terms of specific acceptance criteria for fuel system damage, fuel rod failure, and fuel coolability. Recently, Westinghouse developed the optimized fuel assembly (OFA), which is described in WCAP-9500. WCAP-9500 is ,

mentioned in the last paragraph (p. 4.2-2) of Section 4.2 of the FSAR for Vogtle but is not included in the reference list for that section. WCAP-9500 was approved by the NRC (Tedesco, May 22,1981). The OFA design also consists of a 17 x 17 array of fuel rods but the rods have a diameter of 0.360 in.,

which is somewhat smaller than the rod diameter in the SFA. Because the format of WCAP-9500 followed RG 1.70, some of the fuel design bases and design limits fortheOFAwerenotpresentedinWCAP-9500inaformthatpermittedNRC[o )

cross-check these with the neceptable criteria provided in SRP Section 4.2.

Thertsfore, several questions were issued to clarify the design bases and limits.

Responses to those questions are contained in letters from Westinghouse (Anderson, January 12, 1981, and April 21,1981)[Theseresponsesareapp1T cable to the standard fuel assembly to be used in Vogtle as well (Petrick, September 9, 1981_)./

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) References to these questions and answers will be made at several places in,]the review Q that&follows&. hNYS 4.2.1 De gn Bases "6 4 tr .

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p 4. Y Design bases for the sa t.y analysis address fuel system damage mechanisms and suggest limiting values for important parameters such that damage will be limited to acceptable levels. For convenience, acceptance criteria for these design limits are grouped into three categories in the Standard Review Plan:

(a) fuel system damage criteria,- which are most applicable to normal operation i G plant condition 1), including anticipated operational occurrences LW plant condition 2), (b) fuel rod failure criteria, which apply to normal operation LW 10/20/84 A-3 V0GTLE DSER SEC 4

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the material yield stress and a strain limit of 1%. These limits are unchanged from previously approved Westinghouse fuel designs and remain acceptable for this FSAR.

(9) Assembly Liftoff The Standard Review Plan calls for the fuel assembly holddown capability i

(gravity and springs) to exceed worst-case hydraulic loads for. normal opera- i tion, which includes anticipated operational occurrences. The SFA design basis provides for positive holddown for condition 1, but allows momentary liftoff during one condition 2 event (see Section 4.4.2.6.2 of the FSAR). This design 9

basis is acceptable provided that it can be shown that the affected fuel assem-blies will reseat properly without damage and without other adverse effects during the event. The ability of the affected fuel assemblies to satisfy this provision will be discussed in Section 4.2.3.1, below.

(10) Control Material Leaching .

The Standard Review Plan and General Design Criteria require that control red reactivity be maintained. Control rod reactivity can sometimes be lost by leaching of certain poison materials if the control rod cladding has been breached. The mechanical design basis for the control rods is stated in Section 4.2.1.6 of the FSAR to be consistent with the loading conditions of Section III of the ASME Code. Thus, the design basis for the SFA control rods is to maintain cladding integrity; becaust cladding integrity would ensure that reactivity is maintained, this design basis might appear to be acceptable.

j However, under some circumstances, unexpected breaches might go undetected, so the staff does not normally accept control roo cladding integrity as a sufficient

! design basis.kdiscussion will be presented under Section 4.2.3.1, below, that shows that adequate surveillance will be provided by the applicant to ensure maintenance of reactivity. - ~

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4.2.1.2 Fuel Rod Failure Criteria

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~pW% Q W The evaluation of fuel rod failure thresholds for the failure mechanisms y

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listed in the SRP is presented in the following paragraphs. When these failure 10/20/84 4-11 V0GTLE DSER SEC 4

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These limits are more conservative than the single 280,-cal /gm limit given in RG 1.77. They have been previously approved in the review of WCAP-7588, and i are therefore acceptable.

(3) Cladding Ballooning and Flow Blockage  ;

In the LOCA analyses for SFA-designed plants, empirical models are used to predict the degree of cladding circumferential strain and assembly flow block-age at the time of hot-rod and hot-assembly burst. These models are each expressed as functions of differential pressure across the cladding wall.

There are no specific design limits associated with ballooning and blockage, and the ballooning and blockage models are portions of the ECCS evaluation model, which is documented in Revision 1 of WCAP-9220-P-A and WCAP-9221-A.

(4) Structural Damage from External Forces Section.4.2.3.5 of the FSAR states that the fuel assembly will maintain a

' geometry that is capable of being cooled under the worst-case accident condi-

! tion 4 event and that no interference between control rods and thimble tubes will occur during a safe shutdown earti. quake. This is equivalent to the design basis as presented in the Standard Review Plan and is therefore acceptable.

4.2.2 Description and Design Drawings The description of fuel system components, including fuel rods, bottom and top nozzles, guide and instrument thimbles, grid assemblies, rod cluster control assemblies, burnable poison assemblies, neutron source assemblies, and thimble l plug assemblies, is contained in Section 4.2.2 of the FSAR. In addition, Tables 4.1-1 and 4.3-1 of the FSAR provide numerical values for various core component parameters. While each parameter listed in SRP Subsection 4.2.2 is l not provided in the FSAR, enough information is provided in sufficient detail to provide a reasonably accurate cresantation of the SFA design, and this infa = +4aa 4e thus acceptable. However, the number of fuel rods in D fa Table 4.3-1 of the FSAR should be 50,952 and not 50,592 (see Table 4.1-1 in the FSAR). The applicant should change this value in an amendment to the FSAR. I

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(e) The built-in conservatisms (that is, safety factors of 2 on the stress amplitudes and 20 on the number of cycles) in the strain fatigue analysis as well as the calculated margin to fatigue life limit adequately offset the effect of fretting wear degradation.

Therefore, it is concluded that the SFA fuel rods will perform adequately with respect to fretting wear.

Fretting wear has also been observed on the inner surfaces of guide thimble tubes where the fully withdrawn control rods reside. Significant wear is limited to the relatively scft Zircaloy-4. guide thimble tubes because the f stainless steel control rod claddings are relatively wear resistant. The extent of the wear is both time-dependent and plant-dependent and has, in some non-Westinghouse cases, extended completely through the guide thimble tube wall.

Westinghouse has predicted that an SFA can operate under a rod cluster control Assembly for,m a,pgriod e that exc ds the amo nt of rodded me A-he armou v-u is iMeet in 4ho e imWe

(+exp,ected current 3 scyc schemes b ore fret'f. ng w ar deg on would result in exceeding the present margin to thef g load criterion for the fuel-handling accident. However, the staff requiref several applicants to per- N form a surveillance program because of the uncertainties in predicting wear -

rates for the standard 17 x 17 fuel assembly design. The objective of this program was to demonstrate that g g1 rgmgnroddedguidethimbletubes, thus providing some confidence tha+/ --

'""; is ensured. These applicants DC-formed an owners' group, which has submitted a generic report (Leasburg, March 1, 1982) that provides postirradiation examination results on guide thimble tube wear in the Westinghouse 17 x 17 fuel assembly design. On the basis of this report, the staff has concluded (Rubenstein, April 19, 1982) that the Westinghouse 17 x 17 fuel assembly design is resistant to guide thimble tube wear.

(5) Ovidat 6 and Crud Buildup In the FSAR, there is no explicit discussion of cladcing oxidation, hydriding, and crud buildup. Crud and oxide are mentioned in %ctions 4.4.2.9.1, 4.4.2.11, 4.4.2.11.5, and 4.4.4.5.2 of the FSAR. The applicable models for cladding 10/20/84 4-19 V0GTLE DSER SEC 4 i

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(e) The built-in conservatisms (that is, safety factors of 2 on the stress amplitudes and 20 on the number of cycles) in the strain fatigue analys s i as well as the calculated margin ta fatigue life limit adequately offset the effect of fretting wear degradation.

Therefore, it is concluded that the SFA fuel rods will perform adequately with ,

respect to fretting wear. l Fretting wear has also been observed on the inner surfaces of guide thimble tubes where the fully withdrawn control rods reside. Significant wear is limited to the relatively soft Zircaloy-4 guide thimble tubes because the Inconel or stainless steel control rod claddings are relatively wear resistant. The extent of the wear is both time-dependent and plant-dependent and has, in some non-Westinghouse cases, extended completely through the guide thimble tube wall.

Westinghouse has predicted that an~ SFA can operate under a rod cluster centrol ssembly m riodoftiethatexgeedstheamontofrodded ime R}C A) e 5 with for a,pg$yuelscSe'm eforeNtk w r grNtion rinE32 would result in exceeding the present margin to thef g load criterion for the fuel-handling accident. However, the staff requiref several applicants to per- A form a surveillance program because of the uncertainties in predicting wear rates for the standard 17 x 17 fuel assembly design. The objective of this program was to demonstrate that g gl rm n rodded guide thimble tubes, thus providing some confidence tha p-- 't is ensured. These applicants DC formed an owners' group, which ha bmitted a generic report (Leasburg, March 1, 1982) that provides pos tradiation examination results on guide thimble tube wear in the Westinghouse 17 x 17 fuel assembly design. On the casts of this report, the staff has concluded (Rubenstein, April 19, 1982) that the Westinghouse 17 x 17 fuel assembly design is resistant to guide thimble tube wear.

(5) Oxidation and Crud Buildup l

In the FSAR, there is no explicit discussion of cladding oxidation, hydriding, I and crud buildup. Crud and oxide are mentioned in Sections 4.4.2.9.1, 4.4.2.11, 4.4.2.11.5, and 4.4.4.5.2 of the FSAR. The applicable models for cladding 10/20/84 4-19 V0GTLE OSER SEC 4

3 oxidation and crud butidup are discussed in the supporting documentation (Salvatori, January 4,1973) for the Westinghouse fuel performance code PAD-3.1.

These models were previously approved by the staff. A new temperature-dependent cladding oxGation model is also presented i CAP-91 ecause the f temperature-independ del in pan onservative with respect to 5e2 approved model i WCAP-917 the staff continues to find the older models appli- ,

cable. These models affect the cladding-to-coolant heat transfer coefficient l and the temperature drop across the cladding wall. Mechanical properties and analyses of the cladding are not significantly impacted by oxide and crud buildup. On the basis of the Westinghouse discussion (Anderson, January 12, 1981) of the impact of cladding hydriding on fuel performance, and on our review of the oxidation and crud buildup models, the staff concludes that these effects have been ad'equately accounted for in the standard fuel design.

(6) Rod Bowing l

In Section 4.2.3.1 of the FSAR, the applicant has indicated that the model used fot evaluation of fuel rod bowing is in WCAP-8691 (nonproprietary version is WCAP-8692), which was withdrawn by Westinghouse. Revision 1 to WCAP-8691 was-

" subsequently submitted by Westinghouse and has been approved by the staff f

(Thomas, December 29, 1982). f he applicant needs to use Revision 1 to WCAP-as the reference for the fuel rod bowing model and to confirm that the rod bowing analysis for Vogtle fuel has been performed with this model. This is a confirmatory item. ___

g,9 A S p.b (7) Axial Growth 7 do 4 ,

Relative to the discussion in Section 4.2.1.1(7), above, on stainless steel growth, the staff is aware of supporting information (Bloom, Apiil 1972, and Appleby, April 1972) that was not cited by Westinghouse, but which also implies that irradiation growth of stainless steel should not be significant at the temperatures and fluences that are associated with PWR operation. Furthermore, because the staff is unaware of any operating experience that indicates axial.

growth-related problems in Westinghouse NSSS plants, the staff concludes that 10/20/84 4-20 V0GTLE DSER SEC 4

l from the differential pressure exerted on the fuel rod by the coolant.

l Westinghouse contends that by using prepressurized fuel rods, the rate of clad-l ding creep is reduced, thus delaying the time at which fuel-to-cladding contact ,

l first occurs. The staff agrees that fuel rod prepressurization should improve PCI resistance, albeit in a presently unquantified amount. l 1

In conclusion, Westinghouse has used approved methods to demonstrate that the

' present PCI acceptance criteria have been met. l (6) Cladding Rupture In the LOCA analysis for SFA-designed plants, an empirical model is used to i predict the occurrence of cladding rupture. The rupture model utilized for the large-break analysis is stated in Section 15.6.5.3.1.1 of the FSAR to be the 1981 version of the LOCA evaluation model; however, the references (8 and 13) stated for that model in the last paragraph of that section of the FSAR are incorrect. The correct reference (11, which is Revision 1 of WCAP-9220-P-A

[and WCAP-9221-A] and has been approved by the staff) is in the reference list but is not used 4a the +=vt ction of the FSAR. The applicant should veri -riierences anc mane any ne sary changes in an amendment to the

- fSAR. This is a confirmatory item.

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The rupture model utilized for the small-break analysis was the approved October 1975 version of the ECCS evaluation model (see Section 15.6.5.3.1.2 of the FSAR). This model has been found acceptable for this analysis.

The appropriate references for the large-break LOCA analysis need to be con-firmed. The overall impact of cladding rupture on the response of the SFA design to the loss-of-coolant accident is evaluated in Section 15.6.5 of this report.

4.2.3.3 Fuel Coolability Evaluation The following paragraphs discuss the evaluation of the ability of the SFA fuel to meet the fuel coolability criteria described in Section 4.2.1.3, above.

Those criteria apply to postulated accidents.

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g *' M 8 required to y ASME Code Section XI,1971 Edition through the Winter 1972 Addenda. The applicant has voluntarily updated the preservice inspection program on the basis of ASME Code Section XI,1980 Edition, with addenda through Winter 1980. The use of later referenced Code editions is acceptable as specified by 10 CFR 50.55a(g).

The staff has reviewed the PSI program for the reactor coolant pressure boundary systems and components. As the applicant stated in the FSAR, these systems

. and components are included for examinations per the applicable Code requirements.

The staff established technical positions in the FSAR questions, some of which .

are resolved in the PSI program. The following items require further input or clarification from the applicant:

(1) The PSI program should contain a.. list of components subject to examination and a description of the welds exempted from examination in accordance with IWB-1220 and IWC-1220 and should include the criteria for exemption.

In addition, the examination isometric drawings are necessary for the staff to determine the acceptability of the sample of welds required to ,

beexamined(Q250.1).

(2) In response to FSAR Questions 250.2 and 250.3, the applicant states that ultrasonic examination procedures have been or are being developed for the examination of (a) the reactor coolant pipe and fittings fabricated from SA351, Grade CF8A (centrifugal cast stainless steel) and (b) the reactor vessel which addresses the degree of compliance with RG 1.150 which discusses the near-surface examinations, the resolution with regard [

to detection of actual flaws, and the use of electronic gating as related to the volume of material near the surface of the reactor pressure vessel (RPV) that may not be examined. The staff requests that the applicant provide these examination procedures, so the staff can complete this review.

_QJ (3) ThepreserviceexaminationQreservice)ofthesteamgeneratortubesshould be performed in accordance with NUREG-0452, Revision and RG 1.83, Revision 1, as discussed in Section 5.4.2.gof this SER. M

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be used during subsequent inservice examinations. The staff considers this an g 0, ,

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The Vogtle steam generators are the latest Westinghouse models, designated

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) Model F, which incorporate multiple features to minimize operating problems

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,$ g g { vibration, and dent :'1g. These features include:

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k M %(1) Type 405 ferritic stainless steel quatrefoil tube support' plate to eliminate

, T .s fD denting gh (2) thermally treated Inconel 600 tubing and stress relief of the innermost rows of the tube bundle to reduce the potential for stress corrosion l

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(3) expansion of the tubes to the full depth of the tubesheet to eliminate j crevices and potential for crevice corrosion 0

-j g. (4) a flow baffle plate above the tubesheet to direct lateral flow across the tubesheet surface and thus minimize the number of tubes exposed to sludge and potential for corrosion attack (5) an improved blowdown system to increase blowdown capacity to minimize sludge buildup Vibration-induced wear which has been experienced in the preheater section of Models D and E Westinghouse steam generators, prior to modifications, will not i

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11 RADI0ACUVE WASTE MANAGEMENT n.1 Introduction The~ radioactive waste management systems for Vogtle, Units 1 and 2 (Vogtle) are designed to provide for the controlled handling and treatment of liquid, gaseous and solid wastes. The liquid radioactive waste management system processes wastes from equipment and floor drains, sample wastes, decontamina-tion and laboratory wastes, and chemical regeneration wastes. The gaseous radioactive waste management system provides (1) waste gas decay tanks to allow decay of short-lived noble gases, and (2) treatment of ventilation exhausts through high-efficiency particulate air (HEPA) filters and carbon adsorbers, as necessary, to reduce releases of radioactive materials to as low as is reason-ably achievable (ALARA) levels in accordance with 10 CFR Parts 20 and 50.34a.

The solid radioactive waste mana esent system provides volume reduc ion by drying and incineratio n salidification by using cemen gioTymer binders.

j The radioactive waste hurtragement review area also includes the process and effluent radiological monitoring and sampling system provided for the detection and measurement of radioactive materials in planc process and effluent streams.

l H.1.1 Acceptance Criteria The staff has reviewed the applicant's design, design criteria and design bases for the radioactive waste management systems for Vogtle. The acceptance crite-ria used as the basis for staff evaluation are in SRP Sections H.1,11.2, n.3, 11.4, and 11.5 (NUREG-0800). These acceptance criteria include-the applicable GDC (Appendix A to 10 CFR 50), 10 CFR 20.106, Appendix I to 10 CFR 50, and American National Standards Institute (ANSI) Standard N13.1, " Guide to Sampling  ;

Airborne Radioactive Materials in Nuclear Facilities." Guidelines for imple-

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mentation of the requirements of the acceptance criteria are provided in the  !

ANSI standards, regulatory guides, and other documents identified in SRP Section II. Conformance to the acceptance criteria provides the bases for 10/26/84 11-1 V0GTLE DSER SEC n l

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