ML20214K093

Forwards Rev 1 to Facility Change Safety Analysis ECN R-0832 & Safety Evaluation in Support of once-through Steam Generator Preventative Maint Sleeving Program.Tubes Requiring Repair Will Be Plugged Per Tech Specs

ML20214K093
Person / Time
Site:	Rancho Seco
Issue date:	08/12/1986
From:	Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F Office of Nuclear Reactor Regulation
References
JEW-86-324, NUDOCS 8608180126
Download: ML20214K093 (19)

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,E gsuun SACRAMB.,g NICIPAL UTIUTY DISTRICT P. O. Box 15830. Sacramento CA 95852-1830,(916) 452-3211

• AN ELECTRIC SYSTEM StRVING THE HEART OF CALIFORNIA 15e;'

59-8/ 1 August 12, 1986 JEW 86-324 7

DIRECTOR OF NUCLEAR REACTOR REGULATION ATTENTION: FRANK J. MIRAGLI A, JR. , DIRECTOR j PWR-B DIVISION US NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555

SUBJECT:

OTSG PREVENTATIVE MAINTENANCE SLEEVING PROGRAM j RANCHO SECO NUCLEAR GENERATING STATION t

i

Dear Mr. Miraglia:

As we discussed, we plan to sleeve approximately 250 tubes in the lane and w wedge regions of both the A and B steam generators as a preventative maintenance measure. The intent is to reduce the probability of a primary to secondary tube leak subsequent to the plant restart. The tubes in these areas k (lane and wedge) have demonstrated the highest propensity for developing leaks. None of the tubes to be sleeved as a part of this program degraded beyond Technical Specification limits.

, This application of the sleeving process is not intended to be a repair

process. Snould our current eddy current examinations reveal indications g, requiring repair, we will plug those tubes as required by our Technical Specifications.

I have attached a copy of our ECN R-0832, Rev.1, and the safety evaluation as prescribed by 10 CFR 50.59e for your information. If you have any questions please call me, g Sincer ly, _ _ _ _

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$- Joh . Ward N " Deputy General Manager, Nuclear h

k. Attachment

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(' s F' cc: Syd Miner (Bethesda, MD)

Tony D' Angelo (Rancho Seco NRC) y** -

3 Glen Perez (Rancho Seco NRC) '

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,; - P i ,a 4 DISTRICT HEADQUARTERS 6201 S Street, Sacramento C A 95817-1899 h%.W .%. ( ,

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l Frank J. Miraglia, Jr. August 12, 1986 JEW 86-324 l

bc: W. K. Latham J. E. Ward Chron File R. L. Ashley D. L. Gil11spie G. A. Coward D. Poole S. Knight L. G. Schwieger J. Vinquist D. W. Martin V. C. Lewis (2) R. W. Colombo (2)

C. Andognini P. Turner T. Tucker P. G. Delczenski M. A. Rowden (FFHS&J) Fourth Floor Files Plaza 50 Files Licensing Files .

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i' 7Op SAFETY. REVIEW OF PROPOSED FACILITY CHANGE

1. DESCRIPTION:1 I

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'" ECN 'R-0832 Rev. 1 3 Yes O No O NUMBER OF PAGES /w / attachmenrS):

DESIGN BASIS REPORT REQUIRED:

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2. PRC RECOMMENDATION .

c. Unreviewed Safety Question

• Yes O No s., Change to Facility as Described in SAR  ? Yes 5 D No O

• Yes O No

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• Yes O L No M  ! d. Technical Specification Change

, h Technical Spec:fication Addition . q es ig an aza s Consbradon AnalyMs DISPOS!klON OF PRC

o. Concurs with SA Evaluation and Findings. .H f ,' ' ,

h. 50.54tp) Security Review Required. . . . . . . O

'. i. 50.54(q) Emergency Program Review Required 0  ;

. h Rtcommends Proposal . . . . . . . . . . . . . . . . DD .

, J. 50.54(a) Quality Program Review Required. . .O

c. Sand to MSRC for Ccncurrence . . . . . . . . . QC

d. Fccility Change Rejected . . . . . . . . . . . . . . O t eL Rcturn to Cognizant Engineer. . . . . . ) . . . . O

. Implementing Conditions

f. MSRC Approval Prior to Plant M ork. . . . . . . O .,

g. . Test .of Syr, tem Required . . . . . . . . . . . . . . O RN

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e. Unrsviewed Safety Question Yes O No M h Trchnical Specification Change

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c. Change to Facility as Described in SAR Yes 29 No

' 'c ' DISPb ITION OF NUCLEAR 4' OPERATIONS Yes 29 No MANAGER O '

d. Facility Change Rejected. ...........O ,

a. Plant Work May Proceed ,

. . h Raftr to MSRC . . . . . . . . . . '. . . . . . . . . . . ?I ,

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4. MSRC FINDINGS c j i a. Unreviewed Safety Question , ,

Yes O No O I h T1chnical Specification Change ;

Yes 0 No O c.' Ch ge to Facility as Described in SAR Yes O No O

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. DISPOSITION OF MSRC d. Return to Cognizant Engineer. . . ...O

a. Recommends Change . . . . . . . . . . . . . . . . .O i' h' Send to NRC for Approvel. ........... O  !' e. Facility Change Rejected . . . . . .. . .O I

. c. Plant Work is Not to Proceed . . . . . . . . . . . O

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04TE A s MSRC CHAIRMAN

> 6. ALL WORK REQUESTS COMPLETE AND YELLOW DCNS

[ 5. COMMISSION APPROVAL OSTAINED: , ,4 ISSUED (of applicable):

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' MANAGER, NUCLEAR PLANT DATE 1ECHNICAL SUPPORT SUPERINTENDENT D4rE I 9. DOCUMENTATION COMPLETE: . 10. ADDED TO MONTHLY PEPORT 4 77I r Report Date ouaury u4NAcEn D4rE

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g. nn n . . , , , , , , m . 3, y FACILITY CHANGE SAFETY ANALYSIS RANCHO SECO NUCLEAR GENERATING STATION Log No. 821 Rev. 1 NCH No. WORK REQUEST No.

ECN.No.

R-0832 Rev. 1

Description:

(A description of the desired changes with material and process specifications shall be included.)

See Attached.

• Reason for Change:

(A st::tament as to why the change is being requested.)

See Attached.

Evaluation and Basis for Safety Findings: -

(The evaluation will address itself to specific sections of the SAR or Technical Specifications as applicable. Any effect on nuclear safety.will be described.)

E See Attached.

Safety Findings:

Yes No G O The proposed change is a change to the facility as described in the SAR.

O 3 The proposed change does involve an Unreviewed Safety Question (If an Unreviewed Safety Question is involved, check the appropriate reason).

O Probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously eva!uated in the safety analysis report may be increased.

O Possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis repon .nay be created.

O The margin of safety as defined in the basis for any Technical Specification is reduced.

O G The proposed change does involve a change in the Technica! Specifications.

O 3 The proposed change does involve an addition to the Technical Specifications.

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FACILITY CHANGE SAFETY ANALYSIS ECN R-0832 Rev. 1 LOG NO. 821 Rev. 1

Description:

ECN R-0832 installs approximately 250 Inconel-600 tube sleeves (80 inches long) per steam generator in tubes which are not considered defective. Defective' tubes are defined as tubes which have indications of >40% through wall degradation. The tube sleeves will be installed in accordance with B&W report #BAW-1832P, Rev. 1 "0TSG Mechanical Sleeve Qualification" and B&W supplemental letter 51-1164120-00 "OD Tolerance Justification of 0TSG Mechanical Sleeves" (see ECN R-0832 Rev. 1 and DBR Rev. 1). Defective tubes will be plugged not sleeved.

Reason for Channe:

ECN R-0832 is proposed to permit'the use of steam generator tube sleeving as a preventative measure against OTSG tube failure so as to minimise steam generator tube failure outages. Approximately 250 sleeves per steam generator will be installed in the region of high tube failure probability. This change is an operational enhancement and will improve plant safety by reducing the number of steam generator tube failure outages and also minimise the contamination source of the secondary plant.

Evaluation and Basis inz Safety Findinas:

B&W has issued a report, BAW-1823P Rev. 1,"Once-Through Steam Generator Mechanical Sleeve Qualification", for the repair of degraded OTSG tubes by tube sleeving. The report addresses pertinent aspects of the sleeving process which affect nuclear safety, such as strength and leakage of the mechanical sleeves, corrosion resistance, effect on plant performance, vibration, method of installation, examination for defects and verifications of proper installation, radiological exposure and analysis for transient & accident conditions. The report concludes that the mechanical tube sleeves are qualified for use in degraded OTSG tubes and are strong enough and sufficiently leak-free to significantly exte..d the life of degraded tubes. It also recommends that up to 10,000 of these mechanical sleeves can be installed in the OTSGs to correct or prevent tube degradation.

Tube degradations found in OTSG's have been located most frequently within a few inches above or below the upper tube sheet. Sleeves will be installed in the top portion of the OTSG tubes. Selection criteria for tube sleeving is based on a pattern of tube leaks around the lane region experienced in all operating B&W plants. Defective tubes, when found, will be plugged with a new rolled type plug.

h PAGE 2 0F 3 g .

A comparison of report BAW-1832P and supplemental letter o 51-1164120-00 with the System Design evaluation for the steam L generators in section 4.3.4 of the USAR indicates the design basis for the sleeves is enveloped by the USAR. The report includes an evaluation of the reduction in flow and super heat that would be expected for the worst case in each OTSG and concludes that the effect on plant operation is considered minimal. The B&W report provides sufficient details to conclude the sleeving of up to 10,000 OTSG tubes does not alter the design basis and therefor does not involve an Unreviewed Safety Question.

S Even though the sleeving material used is identical to the original steam generator tube material (Inconel-600) this change does affect the USAR. The design specification of a sleeved tube diameter is less than that of an original steam generator tube

(.525 + .010"/ .002" O.D./.045" minimum wall thickness versus .625

+/- .005" O.D./.034 minimum wall thickness)(See Attachment I, affected USAR page and B&W supplemental letter 51-1164120-00).

o Also, a section should be added to the USAR which discusses tube sleeving and the fact that some tubes are sleeved. Tube sleeving of non-defective tubes for the purpose of preventative maintenance is not affected by Tech Specs.

I A generic analysis performed by B&W on thermal and hydraulic effects assumed 5000 80-inch long sleeves were installed in the j peripheral tubes of each OTSG. This norst case assumption reduced primary flow by less than 1% and reduced steam superheat by 7.7 F at full power. The reduction in superheat results in a 1%

increase in feedwater flow. B&W concludes the slight variations in plant operating conditions would have a minimal effect on B&W plant operation (Reference B&W report BAW-1823P Rev. 1). The District is installing only ~250 sleeves per OTSG (~1/20 the number of sleeves analyzed by B&W).

E This change will sleeve tubes as a preventative measure and will not be used to repair defective tubes as an alternative for tube plugging. Therefore, a change to Tech Specs is not required.

Supplemental Plant Specific Background Information B&W plants were surveyed to determine to what extent other B&W T plants are involved in OTSG tube sleeving activities. Two plants (ANO-1 and Oconee) have either employed or are actively pursuing the use of sleeving for the purpose of repair of defective tubes

' in lieu of plugging. ANO-l's and Oconee's current Tech Specs

- require an amendment in order to sleeve defective tubes.

AP&L sleeved 10 defective steam generator tubes, including 5 with wall perforations, at their ANO unit 1 as a demonstration case for B&W's tube sleeving program. The NRC approved ANO's Tech Spec I allowed the sleeving of the 10 steam generator amendment which tubes. The 10 sleeved tubes at ANO one have performed in an operating reactor one full cycle without any signs of problems.

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AP&L ANO one has had their proposed Tech Spec amendment, which allows tube sleeving for repair of defective tubes, published in the Federal Register.

Duke Power is also seeking a Tech Spec amendment for their three Oconee plants which would allow sleeving of defective tubes for the purpose of repair. Duke Power has submitted their proposed amendment and is presently waiting for e response from the NRC.

Duke Power has responded to questions by the NRC regarding the effects of sleeving on their Oconee Nuclear Station accident analysis. Duke's response uses B&W's generic performance analysis and addresses all events analyzed in their FSAR accident analysis.

Duke Power concluded the plant safety analysis calculations bound --

expected plant conditions with respect to OTSG heat transfer and primary system flow rate, and that up to 5,000 tubes per OTSG may be sleeved without invalidating the existing analysis (See Attachment II, safety review of OTSG tube sleeving). t OTSG tubes, which exhibit some degradation but are not considered defective, are not addressed by Tech Specs other than for surveillance purposes. There is no need to seek a Tech Spec amendment to allow tube sleeving for preventative maintenance purposes, as long as the thermal and hydraulic effects on the plant does not change the accident analysis in the USAR.

B&W's generic performance analysis and Dukes's safety review can 1_

I be equally applied to Rancho Seco. Rancho Seco's design is based on the Oconee plant design. Oconee's accident analysis is similar to Rancho Seco's accident analysis. The attached safety review ~-

addresses the same concerns which sleeving may impose regarding Rancho Seco's accident analysis. Effects of sleeving are minimal and do not alter the accident analysis. The accident 6nalysis for Rancho Seco remains bounding.

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' TABLE 4.2-2 STEAM GENERATOR DESIGN DATA (Data per Steam Generator)

Steam conditions at full load, outlet nozzles Steam flow Steam temperature 6.12 x 10' lb/h 570*F Steam pressure 910 psig Feedwater temperature 470'F Reactor coolant flow Reactor coolant side 68.94 x 10' lb/h Design pressure Design temperature 2,500 psig 650*F Hydrotest pressure 3,125 psig Coolant volume (Hot) 2,030 ft8 Full load temperature inlet / outlet 557.5/606.3'T Secondary side Design pressure Design temperature 1,050 psig Hydrotest pressure 600'T Net volume 1,312.5 psig 3,412 ft8 y< Dimensions g Tubes, OD/ min wall Overall height (including skirt) 0.625/0.03a in.

Shell OD 73-ft-2-1/2 in.

Shell minimum thickness 151-1/8 in.

4.1875 in.

Shell minimum thickness (at tube sheets and feedvater connect) 6.625 in.

Tube sheet thicknesses 24 in.

Dry weight )

1,140,000 lb Tube length (Less length of tube in tubesheet) 52 ft, 1-3/8 in.

Nozzles - reactor coolant side Function No. ID. in. Material inlet 1 36 Outlet 2 Carbon steel - SS Clad 28 Carbon steel - SS Clad Drain 1 1 Sch 160 Manways Inconel 2 18 Handholes 2 Carbon steel - SS Clad 5

Carbon steel - SS Clad Netzles - secondary side Steam 2 24 Vent Carbon steel 1

1 1/2 Sch 80 Carbon steel N A/c' le .sfee vect n4e flow reshn J is e e.s- h /c n' at 3 h % ./. O.045 thh 4.2-5

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Ud NIIN ENGINEERING INFORMATION RECORD a McDermott comparty Safety Related:

Document Identifier 51 1164120-00 Yes O No O w ..

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Title O h ERANCE JUSTIFIL IION - OTSG MECHANICAL SLEEVE ~

Prepared by *> Date 6 i

Reviewed by -

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1. Date N Mf, 7 JC Remarks:

The desired sleeve outside diameter was originally specified aa .525" t .002; this dimension was specified in the B&W Sleeve cQualification Report (BAW-1823P) J and in the purchase order for sleeve tubing. The tight OD tolerance (2.002) "

was originally specified because sensitivity of the curving tool to sleeve diameter variation had not been established.

In subsequent sleeve fabrication and material procurement, it has been found that the OD tolerance is too restrictive from a practical standpoint. In the worst case, some material OD was found to actually be .525" +.010/ .002.

Prior to fabricating sleeves, samples of the worst case material were curved with the curving tool and then inserted in the SPIS OTSG mockup. This confirmed the acceptability of .525 +.010/ .002 material for curving and installation.

Prior to installation, each actual measured sleeve OD and ID are input to an equation which in turn sets the roll expander for the optimum mechanical joint. Therefore the joint expansion controls account for the actual CD of each sleeve. Since the joint expansion can be controlled within the desired range, joint integrity is assured for any sleeve OD.

Sleeves with larger OD's should be less susceptible to corrosion since the material is displaced less during expansion before it contacts the parent tube than sleeves with smaller OD's.

Based upon the above, Steam Generator Engineering concludes tnat

.525 +.010/ .002 OD sleeves can be curved and installed and that mechanical joint integrity will be acceptable. It is further concluded that these sleeves will be no less corrosion resistant than those originally specified and that all qualification tests summarized in BAW 1823P are valid and applicable.

Page / of /

1 ..

ATTACHMENT Safety Review of Steam Generator Tube Sleeving for Oconee Nuclear Station Thio report summarizes a review of the Oconee Nuclear Station accident analysis to document the impact on plant safety of slesving five thousand (5,000) tubes par steam generator.

Stoism generator slesving has been shown to be an effective means of ropciring degraded steam generator tubes. By aleaving the tube rather than plugging, the tube remains in service and continues to be used to transfer enorgy from the primary to secondary system. The sleeved tube perforns as well mechanically as an unsleeved tube and does not significantly increase the probability or consequences of accidents previously analyzed and does ust create the chance for a new event that is nor already bounded by the liconcing analysis. However, the sleeved tube vill not transfer energy as officiently as an unsleeved tube. Thus, the sleeving will result in a slight reduction in heat transfer in the steam generator and a small incrase in the primary side pressure drop through the steam generator due to tha smaller tube diameters in the sleeved tubes.

Ao discussed in Reference 1, the effect of sieering 5,000 tubes in one gen 2rntor is a reduction in primary flow of less than one percent and a decrocse in steam auperheat of approximately 7.7 F at full power. The rsduction in superheat results in the need for an additional 1% full fandvater flow to remove the same amount of primary energy.

Th3 FSAR analysis of overcooling events assumed that the feedwater flow inerecsed during the event to increase the heat removal by the steam gancrator. Thus, the snail increase in nominal feed / steam flow does not impact the safety analysis of the Oconee units since the heat removal rates cciculated for chase events are conservative.

Fcr everheating events, the heat transfer in the steam generator is either left at the nominal value, which will not change due to tube sleeving, or is reduced dramatically for events such as a loss of main feedwater. The slight reduction in heat transfer coefficient along the sleeve will not impact these assumptions since the plant is in a steady-state heat transfer candition prior to the event.

Other events analysed in the FSAR which do not fall into one of the above categories are not affected at all since the assumption on steam generator heat removal does not change.

Ths effectiveness of AFW cooling vill be decreased somewhat due to the incertion of sleeves. The effect will be that the heat removal vill occur I at a lower elevation in the steam generator, thus slightly lowering the thornal centar of the generator. This will not significantly affect the chility of the generator to operate in natural circulation or boiler condenser modes of cooling.

-z-m e 2 d'2 A dderooso in primary cyntos flow haa boon ovaluated previously for DtI-1 undar a task to evalusta plugging up to a total of 3.000 steam generator tubas. The evaluation summarised in Referenea 2, aoncluded that the primary system flow would be reduc,ed by approximately 2.5%. 'Ihis reduction is larger than that for the tube sleeving under review in this instance.

Tho conclusions of the TMI-1 review indiested sufficient design margina exist to a: low full power operation with the large number of tubes plugged.

Howrover, to assure that the plant safety analysis remains bounding, it is required to measure primary . system flow at the begfaadag of each fuel cycle and demonstrate that the act;ual flow is in excess of the flow rata assumed in the analysia. As long 'as the minimum flow rate is net, the plant safety analysis remains valid. Duke Power will measure primary flow each fuel cycle to ensure the minimum flow used in design calculations exists. This i vorification will ensure that the safety analysis calculations for the Oconce units remains bounding. .

In conclusion, since the piant safety analysis calculations bound the expasted plant conditions with respect to stasa generator heat transfer and prima ~ry system flow rate, up to 5,000 tubes per steam generator may be I cleaved at the Oconea units without invalidating tha existing analysis bcGois of the units. ,

Refcronces:

(1) BAW-1832P Ray. 1 "Oncerthrough Stesa Generator Nachanical 51 eave Qualification", Babcock,& Wilcox, Lynchburg, Virginia ' November, 1985.

(2) B&W. Document Number 86-1130560-00 "Eiket of Plugging OTSG Tubes".

Babcock & Wilcox. Lynchburg Virginia. Janc ry 1982.

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i hMQQ sAcnAMENTO MUNICIPAL UTtuTY DISTRICT ECN NO. R-0832 REV 1 Sheet Of

' ENGINEERING CHANGE NOTICE -m. 2 1 c 3r a cus.

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.n O THE FOLLOWING SYSTEM (S) WILL BE AFFECTED. SY THIS CHANGE

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m......m=.m-,,.=... M, woan oaosa ac =ca =c O REASON FOR CHANGE 104499 Preventive sleeving of steam generator tubes Tv_*E, 05 Ecx simouLAn I _ MAJOR _~ $US 2 D DESCRIPTION OF THIS CHANGE PROVICE A DETABLED DESCRIPTION ANO ITS EFFECT ON PLANT OPERATION. INCLUDE SKETCHES AND OTHER PENTINENT INFORMATION TO COMPLETELY DESCRest ANO nLLUSTRATE MOW CHANGE IS TO FUNCTION AS WELL AS ILLUSTRATE INTERFACEIS) WITM EXISTING EQUtPMENT INDICATE WHAT NEW COMPONENT!St ARE REQUIRED. TYPE S O F O R AWINE S AFP ECT E D A N D E STt M ATED TOT A L CO ST OF PRO J E CT.

L 8RIEF TIT 1.E OF ECN:

OTSG sleeving.

2. DESCRIPTION OF CHANGE: -

Install

.045 min approximately wall) per unit in 250 both Inconel-600 steam generators tube sleeves E-205 A&B. (.525 + .010/ .002 inODby[^l\

The sleeve design and quality requirements are in accordance with B&W document BAW-1823P, Rev. 1 "0TSG Mechanical Sleeve Qualification" A and supplement 51-1164120-00 "0D Tolerance Justification of OTSG /l\

Mechanical Sleeves."

Work under this ECN is a preventive measure with sleeves installed in non-defective tubes only. No sleeves will be installed in defective tubes. (Defective tube is defined in Tech Spec as a tube with more than 40% wall degradation).

Affected drawings: N6.03-46 Sht 1&2 Total estimated cost: $5,750,000 N6.03-81, Sht. 2&3

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DESIGN B ASIS RE PORT com esv me. I carg Atmust 6th.1986 oisc PLINE TMis oSA IS B ASED QN N C R N o. WORK REQUEST Ne N#C "" R-0832 Rev.1 104499

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1. PURPOSE OF QESIGN CH ANGE:

See Attached

11. DESIGN CRITERI A USED:

l See Attached 111. CALCUL ATIONS & DESIGN IN FOR M ATION:

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. IV. FAILUR E MO DES:

$ Taas cManet moss mor arracT comTaot noons insinumanTarion O Tais cuaaes arrects coaTaoL acou insTavanafation. sus amaLTsis V. SPECI AL M AINTENANCE REQUIREMENTS:

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V l. SPECI AL OPER ATING REQUIR EMENTS:

None ,

Vll. VERIFIC ATION CRITE RI A :

See Attached Vill. COMMENTS : .

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so Design Basis Report ECN R-0832 I. Purpose of Design Change:

The purpose of this design change is to install Inconel 600 tube sleeves in non-defective steam generator tubes. The sleeving of the tubes is a preventive measure devised to mitigate the spread of steam generator tube leakages. Approximately 250 tubes per generator will be provided with Inconel 600 sleeves in the region of high tube leakage probability.

II. Design Criteria Used:

A. Sumary of Change This modification consists of selecting about 250 tubes in the leak prone region of the steam generator and installing sleeves in them. The selection criteria is based on a pattern of tube leaks experienced in the operating B&W plants. Only non-defective tubes will be sleeved. Defective tubes, when found, will be plugged with a new rolled type plug.

B. Design Bases

1. The sleeve design is in accordance with B&W document BAW-1823P, Rev. 1 and supplement 51-1164120-00

2. The rolled plug design is in accordance with B&W document -

BWNP-20440-2 C. Scope This change includes identifying OTSG tubes for sleeving and plugging with the new rolled type plug given the initial 250 selected tubes. The sleeving and plugging operation will be in accordance with installation procedure M-44, "0TSG Tube Sleeving".

D. Equipment Class and Power Requirements Equipment installed under this modification bears Project l .'!,\ '

Classification 11. There are no power requirements for this new equipment.

E. Testing The installation procedure M-44 contains the requirements for baseline Eddy Current testing which will verify the acceptability of the installed sleeve.

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. ' Design Basis Report ECN R-0832 Page 2 of 2 III. Calculation and Design Information:

A. Design Features The OTSG sleeves are made from Inconel 600 having

.525 +.010/ .002 IN 0.D. by .045 wall. The sleeves also act A as stabilizers. /I\

B. Functional Description The function of the sleeves is described in the B&W document BAW-1823P "Once-Through Steam Generator Mechanical Sleeve Qualification" and supplement 51-1164120-00 "0D Tolerance Justification of OTSG Mechanical Sleeves".

C. Design Calculations None IV. Failure Modes:

A. Sleeving is done to preclude OTSG primary to secondary leakage. Should the sleeved tubes leak, the secondary side is i monitored and plant will shut down when the secondary side approaches Technical Specification limits.

V. Special Maintenance Requirements:

None VI. Special Operating Requirements:

None VII. Verification Criteria:

A. Verification of the expansion of the bottom joint is done by measuring the I.D. of the sleeve, Eddy Current testing and comparing it with values contained in the installation procedure M-44.

B. Eddy Current testing of the completely installed sleeve to verify that no defect has been incurred during installation.

VIII. Comments:

None 039mb

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RANCHO SECO NUCLEAR GENERATING STATION l DESIGN VERIFICATION REPORT MOD No. Nfk ECN No.(-DW i k ! RELATED ECN'S b l

METHOD: O QUAUFICATION TESTING

%]INTERCISCIPUNARY INDIVIDUAL REVIEW REVIEW O ALTERNATE CALCULATION DOCUMENT (S) REVIEWED:

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SUMMARY

OF REVIEW (Attach additxvlal sheetts) if needed):

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CON USIONS: (Attach addinonal sheet (s) if needed) b i v -. dir.o vWC

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a SHET 2 OF 4 DESIGN VERIFICATION REPORT BASIC CHECK POINTS Reviewer's

- N/A initials .

A. 19 Basic ct@.6 from ANSI 45.2.11-1974, Section 6.3

1. Design inputs were correcdy selected and incor-porated in the design. (Ref. ANSI 45.2.11, Sec-tion 3.2) g

2. Assuir,As.s necessary to @fw. i. the demgn ac-tivity are adequately desenbod and reasonable.

Also where necessary, design assumptions are identified for subsequent revenfication when the detailed desgn activities are completed. -

3. The appropriate quality and quality assurance re-quirements are specified.

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4. The appiir shia codes, standards and reguistory re-quirements, including issue and sddenda, are pro-periy idennfled, and their requirements for desgn have been met p

5. Applicable construccon and openmng exponenca g have been considered.

6. Design interface requirements have been sansfied. U Appropnate design trethod has been used. OLLA ~ -

7.

8. The design output is reasonable compared to the input. IluA-

9. The specified parts, equipmentand processes are suitable for the required application. I

10. The specified matenais are co..wm with each -

other and with the design environmental condi-tions to which the matenal will be avra=ari.

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11. Adequate maintenance features and requirements are specified. ,

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- SHEET 3 oF 4

- Reviewer's N/A initiais

12. Accessibility and other provisions for maintenance i and repair have been included. l

13. Accessibility for expected ISI during plant life has been included.

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14. The design has procerty considered radiation ex-posure to the public and plant personne!. ALARA p%'

program requirements have been met.

15. Adequate acceptance enteria have been included in design documents to allow venfication that design requirements have been satisfactorily accomplished.

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16. Adequate pre-operational and subsequent penodic test requirements have been specified where applicable.

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17. Adequate handling, storage, cleaning and shipp- _

ing requirements have been specified (Also see NEP 4206 " Purchase Request").

18. Adequate identificanon requrements are sp.dGod. /1 -

'19. Requirements for record preparation, review, ap- y proval and rotermon are adequately specified.

B. The foNowing ctM.;. are Rancho Seco specific.

1. HELBA program requirements have been met.

A/A-

2. Fire protecton and FHAR have been adequataety j considered. Nik -

3. The demgn has been reviewed for Human Factors enginsonng.

4 4. Piccding wtontimi has been adequately conadered.

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5. Impact on IDADS has been considered, includmg software changes. Oglk

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6. Seismic Il-over-seismic I design has been considered. N,F c

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• SHEET 4 OF 4

-- Reviewer's N/A initials

7. RG 1.97 impact reviewed. Nb

8. NUREG 0737 commitment imoact reviewed. N,lk-

9. SPOS data base impact reviewed. OL

10. Impact on Emergency Response Plan has been considered. k V

11. Impact on Emergency Response Capabilices has been considered. .

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12. EQ Program Requirements have been sansfied.

k C. Addidonalitems Reviewed --

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D. fApprovals

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