ML20038A749
| ML20038A749 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 11/09/1981 |
| From: | Hood D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8111160369 | |
| Download: ML20038A749 (39) | |
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i N' OV 0 1931' Docket Nos: 50-329 and 50-330 APPLICANT: Consumers Power Company FACILITY:
Hidland Plant, Units 1 and 2
SUBJECT:
SUlftARY OF SEPTEliBER 21, 1981 MEETING ON FUEL DESIGH On September 21, 1981, the llRC staff met in Bethesda, liaryland, with Consumers Power Corpany (the Applicant), and Babcock A Wilcox. The discussions included conformance of Midland FSAR Section 4.2, " Fuel Design" to Revision 2 of that section as given in the Standard Review Plant (SRP), NUREG-0800, July 1981. The Midland FSAR was pre-parad to Revision 3 of Regulatory Guide 1.70, which corresponds, to Revision 0 of SRP Section 4.2.
The discussion also included the liidland approach to several technical issues which had previously been outstanding on other plants reviewed by the staff.
A sunnary of these discussions, including copies of the visual aids and handouts used during presentations, and identification of meeting attendees, are enclosed.
Darl S. Hood, Project Manager Licensing Branch flo. 4 Division of Licensing
Enclosures:
As stated cc w/ enc 1:
See next page 8111160369 811109 PDR ADOCK 05000329
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MIDLAND Mr. J,1 Cook Vice President Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201 cc:
Michael I. Miller, Esq.
Mr. Don van Farrowe, Chief Ronald G. Zamarin, Esq.
Division of Radiological Health Alan S. Farnell, Esq.
Department of Public Health Isham, Lincoln & Beale Pc0. Box 33035 Suite -4200 Lansing, Michigan 48909 1 First National Plaza
- Chicago, Illinois 60603 William J. Scanlon, Esq.
2034 Pauline Boulevard James E. Brunner, Esq.
Ann Arbor, Michigan 48103 Consumers Power Company 212 West Michigan Avenue U.S. Nuclear Regulatory Commission Jackson, Michigan 49201 Resident Inspectors Office Route 7 Myron M. Cherry, Esq.
Midland, Michigan 48640 1 IBM Flaza Chicago, Illinois 60611 Ms. Barbara Stamiris 5795 N. River Ms. Mary Sinclair Freeland, Michigan 48623 5711 Summerset Drive Midland, Michigan 48640 Mr. Paul A. Perry, Secretary Consumers Power Company Stewart H. Freeman 212 W. Michigan Avenue Assistant Attorney General Jackson, Michigan 49201 State of Michigan Environmental Protection Division Mr. Walt Apley 720 Law Building c/o Mr. Max Clausen Lansing, Michigan 48913 Battelle Pacific North West Labs (PNWL)
Battelle Blvd.
Mr. Wendell Marshall SIGMA IV Building Route 10 Richland, Washington 99352 Midland, Michigan 48640 Mr. I. Charak, Manager Mr. Steve Gadler NRC Assistance Project 2120 Carter Avenue Argonne National Laboratory St. Paul, Minnesota 55108 9700 South Cass Avenue Argonne, Illinois 60439 Mr. Roger W. Huston Suite 220 Mr. R. B. Borsum 7910 Woodmont Avenue Nuclear Power Generation D1 vision Bethesda, Maryland 20814 Babcock & Wilcox 7910 Woodmont Avente, Suite 220 Bethesda, Marylant 20814 L1
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..e itEETIN 541 NARY - HIDLAND FSAR SECTIGM 4.2 SEPTD'aER 21, 1981 Attendees:
HRC CP Co B&W D. Hood, Project !!anager L. S. Gibson F. Levandoski R. O. Heyer, CPB T. C Hollowell W. Gray f,. Powers, CPD R. W, Huston G.11 eyer H. Tokar, CPD (Parttime)
R. Demars R. Gupta, CPB (Parttine)
J. Tulenko W. Jensen, RSB (Parttime)
H. Bailey R. Voglewede, CPB (Parttime)
J. Shah Meeting
Purpose:
To discuss the differences between Section 4.2 of the Midland FSAR and Revision 2 to Section 4.2 as oiven in the July 1981 SRP (WUREG-08CD). Also to discuss several recent technicu issues cited as open issues in recent SERs on other plants.
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The meeting agenda is attached. All items on the agenta were discussed alt. hough the "Recent Technical Issues" were covered in a different order than shown.
The attached Figures 1-30 (overhead slides) were presented by representatives of Consumers Power Company and Babcock & Wilcox.
The meet % began by describing the differences between the FSAR Standard Format (Reg Guide 1.7C; and the Standard Review Plan. Hidland FbAR Section 4.2 was prepared in accordance with Reg Guide 1.70 (Rev 3) while it is being reviewed by the NRC's Core Performance Branch (CPD) based on the requireocnts of Revision 2 to Section 4.2_ as given in HUREG-0800. The primary difference in the two documents is the format and the SRP_ requirement for more quantitative design bases, limits and analyses. Consumers Power presented a cross-reference guide, Figures 3, 4 and 5 which indicate the sections of the FSAR where many design bases requirentnts of the SRP are located. There are some areas which were inadequately covered in the FSAR. These cases were indicated with the
. statement " Supplementary Information Requirved." The design bases which were inadequately covered included fretting wear, oxidation and crud buildup, and pressure buildup for hurnable poison rods.
In addition to the review of FSAR Section 4.2 for_ compliance with the Standard Review Plan, Consumers Power was also asked to discuss seven "Recent Technical Issues" which have been cited in recent SERs. These issues are. delineated in Figure 6. -It was pointed out that sone of these issues were considered resolved for the Midland' Plant.
Two notable exceptions, which were discussed in detail later,'were supplemental ECCS I
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% analysis with HUREG-630 and combined seismic and LOCA loads analysis.
In both of these case'. the applic, nt expressed its' belief that final or supplemental. calcula-tions yet to be performad would confirm safe operation of the plant as designed.
'The CPB reviewer psked if all core components were included in Section 4.2 of the
,FSAR. The app 1Laat replied that only fuel, control and burnable poison components, including all structural components, are coverM in Section 4.2.
In-core instru-ments (movable type) are covered in a different section of the FSAR.
(A post-meeting check has indicated that the novable in-core instruments are covered in Section 7.8.2 of the FSAR.) CPD will request additional information on neutron source rods and other cogonents as needed to comlete its review.
Various staff nenbers from Babcock & Wilcox (B&W) presented more detailed information to supplement the FSAR. This began with a presentation on fretting wear (Figures 7 -
10).
BSW indicated that the fatigue analysis uses a stress concentration factor of 2.0 to cover all types of cladding imperfections of which fuel rod fretting is classed as one type. Tests and post irradiation examinations have shown fretting to be localized
- ar.d less than 107, of the cladding wall. The B&W conclusion is that the analyses per-forned adequately account for fretting and show a large positive margin. Guide tube wear was also addressed under fretting wear.
ine NRC representatives indicated that the BAW topical report on this subject (DAW 1623) had been approved in July 1981. The conclusion was that guide tube wear will not be an important limiting consideration for the Midland Plant. The CPB reviewer asked if control rod finger wear has been considered where the control rods enter the fuel assembly. B&W replied that control rod wear, other than surface scratches, has not been observed in B&W plants. The reason is believed to be related to the fact that the control fingers are continuously enclosed in the region above the fuel assembly. This eliminates vibration producing cross flow forces on the control rod fingers and also eliminates a transition where the fingers enter the fuel assembly.
Cladding oxidation and crud behavior were discussed, first from a mechanical stand-point and then fron a thermal effects standpoint (Figures > 1-17). The data available fron PIE on B&W fuel indicates the maxinum cladding oxtetton expected at the target burnup of 37,000 INd/MTV should be less than 1 nil. No acceleration of oxidation with
.burnup has been observed by B&W. llechanical analyses of cladding account for the expected oxidation by assuming a wall thickness reduction from nouinal which is 2 mils whereas the tolerance on wall thickness only allows 1 mil reduction. Crud deposition observed by D&W has been greatest after one cycle but has not exceeded 1 mil. Crud DFF6CE )
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. thickness is observed to diminish with burnup. The B&W thermal model (TAC 0, TACO-2) does not explicitly treat oxide thickness and crud buildup in the Hidland Plant analysis..These factors are implicitly included in the nodel however, because bench-marking has been done against real fuel rod performance data where oxidation and crud buildup effects have been reflected in the measured temperatures. TACO-2 has the
. capability to treat an oxide layer explicitly. B&W stated that scoping calculations assuming i nil oxide layer and 1 mii crud indicate approximately 27 F total delta T could exist across these layers. This delta T is assumed to exist in the experimental rods utilized for benchmarking of TACO and TACO-2. TACO-2 conservatively predicts the benchmarking data.
The design basis for internal pressure in burnable poison rods is that the internal pressure should not exceed system pressure during the residence time in the core.
~ The applicant finds that this criterion has been met by its analysis.
The first of the "Recent Technical Issues" to be covered was the fission gas release at high burnup and the end-of-life fuel rod internal pressure analysis. The Midland fission gas release / internal rod pressure analysis utilized the TACO mooel with the NRC burnup enhancement model applied. This approach is acceptable to the CPB. The TACO-2 model, which is currently' under NRC review, includes a new temperature-burnup dependent fission gas release model.
Fuel rod bow has been analyzed for Midland using the NRC. rod bow interim criteria.
The DHBR penalty due to rod bow is partially offset by the 1% pitch reduction factor.
WitP offset applied there ray be no net DNBR penalty for Hidland Core 1.
Under this interin method, the net DNBR penalty at 40,000 INd/HTU increases to 5.07,.
The basis of the technical specifications, when they are finalized, will include whichever DHDR penalty nethod is approved by the HRC at that time. The CPB reviewer stated that the DNBR penalty finally applied should be based on the lead burnup assembly rather than the limiting (power wise) assently which-is normally at a lower bu rnup. The applicant stated that its response on this analysis will be supplemented.
The fuel assembly design shoulder gap analysis was briefly reviewed (Figure 25). This analysis indicates the design basis of no axial interference during the life of'the fuel will be met.
The appitcant has elected to delay the ECCS analysis (supplementary) for Midland utilizing the clad swelling and rupture models in NUREG-0630 to allow the incorpor-ation of other new nodels in the analysis at the same time. The TACO-2 model will be one part of this updated program. The CPB is currently reviewing the TACO-2 model with acceptance anticipated within a very short time. Other required nodels to be reviewed by the NRC -include Theta 1-B and REFL000-3. The approval date of these codes onice >
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. is unknown.. As part of the incorporation of NUREG-0630 it will also be necessary to revise and resubmit CRAFT-2 and Theta 1-B with the new swelling and rupture (see Figure 27) models. Consumers Power indicated it would not be cost effective to rerun the ECCS analyses with NUREG-0630 without incorporating all the other model changes
.at the same time.
Consumers Power Company will submit a plan of action to the NRC indicating how it pro-
. poses to proceed on supplemental ECCS analyses, what schedule will be followed, and what NRC approvals of other models (revisions only) will be required and the dates required. The proposed approach will include other subnodel tradcoffs to be applied to offset the effects of the NUREG-0630 implenentation. Similar offsetting model changes have been accepted by the NRC for recent C-E and Westinghouse supplementary ECCS analyses with NUREG-0630.
The applicant stated that it has a high level of confidence that the final ECCS analyses, incorporating huREG-0630 and other model changes, will confirm the existence of similar or possibly lower peak cladding temperatures and peak cladding oxidation comared to the previous LOCA analyses.
The seismic /LOCA loading analyses for the Hidland core also require additional cal-culations. The pressure vessel upper lateral support addition, and revisions in the earthquake loadings are the reason. The original Hidland seismic analyses were covered by BAU topical report BAW 10035. New analyses addressing asymmetric LOCA, t
combined SSE and LOCA, and Midland unique characteristics will be performed. Topical report DAW 10133P, which describes the B&W evaluation model, was submitted in May 1979 and responses to questions were submitted in July 1980. The responses to questions essentially meet the requirements of NUREG-0800. Asymetric LOCA loading in 177 fuel-assembly plants was covered in BAW 1621.
The NRC review of the DAW topical reports is being accomplished by an outside con-tractor. The schedule for cogletion of the reviews was not known at the time of the meeting, but NRC will advise Consumers Power and B&W of the status of these reviews.
The revised final seismic /LOCA analysis for the Midland core will be completed within six months of definition of the input seismic loadings. This definition is-currently expected by May 1982. The applicant expressed confidence that the final Hidland analysis will show positive margin (see Figure 30). This confidence results from a cogarison with other plant analyses where the setsnic loadings have been significantly higher than expected for the Hidland Plant.
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The CPB indicated that the FSAR cross-referencing approach (Figures 3, 4 and 5 attached) was satisfactory for providing auch of the information required by the Standard Review Plan. When the review begins in detail there will be some additional
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clarification required. The NRC/CPB reviewer and the Consumers Power Co:npany repre-sentatives will maintain close contact to resolve all clarification questions as expeditiously as possible.
The schedule for the Midland SER was discussed. The CPB will meet the liay date for the final SER. Because of other priorities, however, the detailed review and draft SER preparation has not yet begun. The draft SER will be available during the first quarter of 1982.
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Proposed Agenda for Meeting With Core Performance Branch September 21, 1981 I.
Introduction (CP Co)
A.
Discussion of Requirements of SRP vs Standard Format,'specifically Section 4.2.
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B.
Cross-reference chart locating SRP (4.2) required information in Midland FSAR.
C.
Delineation of insufficiently covered or missing items required by SRP II.
Detailed Discussion of Insufficiently Covered or Missing Information Required by SRP (B&W)
III.
Discussion of "Recent Technical Issues" Cited as Outstanding Issues in Recent SERs.
Presentation of B&W/CP Co/ Resolutions of These Issues as They Apply to Midland (B&W) frapese d,
h A.
Supplemental ECCS Analysis with NUREG-0630 j)
B.
Combined Seismic /LOCA Loads Analysis (f)
C.
Enhanced Fission Cas Release Analysis at High Burnups (2)
D.' Fuel Rod Bowing Analysis g/Ldyj((
E.
Guide Tube Wear Analysis
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Shoulder Cap Analysis (h
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End-of-Lif e Fuel Rod Internal Pressure Analysis IV.
Wrap-up Discussion - Agreement on any further information which may be required by Core Performance Branch to co=plete Midland OL review.
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' REVIEW OF MIDLAND FSAR SECTION 4.2. -
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.FOR COMPLIANCE WITH'THE STANDARD REVIEW PLAN
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. ac 1 COINMONALITY BETWEEN STANDARD FORMAT AND STANDARD REVIEW PLAN
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(SECTION 4.2)
Four Areas of Information/ Review
- 1. DESIGN B'ASES I
- 2. DESCRIPTION & DESIGN DRAWINGS
- 3. DESIGN EVALUATION
- 4. TESTING, INSPECTION, AND SURVEILLANCE PLANS I
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MOSTIMPOR NT DIFFERENCES ARE IN THE AREA OF DESIGN BASES Standard Format nu.3(sues,4u,u-,,,a,,.zyf,,.g 4.2.1 DESIGA BASES
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- 1. Cladding
- 2. Fuel Material
- 3. Fuel Rod Performance l
- 4. Spacer Grid.and Channel Boxes
- 5. Fuel Assembly
- 6. Reactivity Control Assembly -and Burnable Poison Rods
- 7. Surveillance Program Standard Review Plan (R" =#"^
"5 11 ACCEPTANCE CRITERIA A. DESIGN BASES
- 1. Fuel System Damage
- 2. Fuel Rod Failure
- 3. Fuel Coolability I
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1 CROSSREFERENCEGUIDE-SRPSECTION4'2(Bev.Z)
SRP Design Bases Where Information is
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Requirement Provided _ in FSAR - " " ~- --
II.A.1 Fuel System Damage Stress, Strain or 4.2.1.5.1 Structural Design Loading Limits B.
Strain Fatigue 4.2.1.1.3 Vibration and Fatigue z
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Fretting Wear
. 4'.2.3'.1.4 Fretting & Crevice' Corrosion, Also
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BAW 1623 Guide Tube Wear. Topical
-Supplementary Information_ Required D.
Oxidation, Supplementary Information Required Hydriding, (Oxidation & Crud)
& Crud Buildup'
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'Hydsiding
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Dimensional Changes 4.2".3.1 8
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4.2.3.1.10 Irradiation Stability 4 2.3.4.1
- Dimensional Stability 4.2.3.5.1 Fuel Assembly Loads F.
Fuel & Burnable 4.2.3.1.2 Internal Fuel Pin Pressure Poison Rod Supplementary Information Internal Pressures Required for BP Rods G.
Hydraulic Loads 4.2.3.5.1 Fuel Assembly Loads (Holddown)
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Control Road 4.2.3.6.1 Rod Pressures and Cladding Reactivity Stresses Maintenance 4.2.3.6.4 Potential for Chemical Interaction 4
miO981-0278a131 A
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SRP Design Bases Where Information is Requirement Provided in FSAR II.A.2 Fuel Rod Failure
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A.
Overheating 4.2.3'.1.1 (b) Fuel Thermal Analysis 4.2.1.5.2 Thermal-Hydraulic Design B.
Pellet / Cladding 4.2.3.1.3 Potential For Chemical Reaction
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Interaction 4.2.3.3.1 Pellet Cladding Mechanical Interaction C.
Hydriding 4.2.1.1 And Proprietary Submittal in Response to Question 231.26 D.
Cladding Collapse 4.2.1.3.1.a Creep Collapse Analysis E.
Bursting 6.3.3.1 ECCS Evaluation Model Also Topicals BAW 10104 BAW 10103 F.
Mechanical 4.2.3.5.2 LOCA and/or Seismic Loading Fracturing G.
Fretting Same as II. A.1.c - M :g.
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FIG 5 CROSS REFERENCE GUIDE - SRP SECTION 4.2 { M -
SRP Design Bases Where Information is Requirement Provided in FSAR II.A.3 Fuel Cholability A.
Cladding 6.3.3.1 ECCS Evaluation Model Embrittlement-4.2.3.3.6 Energy Release From Cladding -
Coolant Reaction B.
Vicient Expuls' ion j4.2.3.3.4. _ Potential Effect offFuel Rod of Fuel Rupture 4.2.3.3.7 Energy Release and Resulting Pressure Pulse From Waterlogged Rupture
- 15.4.8 Rod Ejection Accident C.. Generalized..
. ;. Covered by A'.
Above Cladding Melting; D.
Structural
,_. 4.2.3.5.7 LOCA and/or Seismic Loading Deformation E.
Fuel Rod Covered by A. above Ballooning e
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l RECENT TECHNICAL ISSUES CITED l
- 1. Supplemental ECCS Analysis with NUREG 0630
- 2. Combined SEISMIC and-LOCA ~
Loads Analysis
- 3. Enhanced Fission ~ Gas Release Analysis at High Burnup
- 4. Fuel Rod Bowing Analysis
- 5. Fuel Assembly Control Rod Guide Tube Wear Analysis
- 6. Fuel Assembly Design Shoulder Gap Analysis
- 7. End-of-life Fuel Rod Internal Pressure Analysis e
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FRETTING WEAR
'hESIGNNASIS FUEL RODS SHALL NOT Fall DUE TO EXCESSIVE FRETTING DURING. CONDITION I & 1(EVENTS.:
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v DESIGN EVALUATION TESTS AND PIE DATA SHOW FRETTING TO BE.._.
LOCAllZED AND LESS THAN.10% OF WALL THICKNESS (SEE FSAR ll.2',3,1,10.
FATIGUE ANALYSIS USES STRESS CONCENTRATION FACTOR-OF 2.0 TO COVER CLADDitlG IMPERFECT 10ilS 9
(IN ADDITION TO 2 AllD 20 TREATHENT OF FATIGUE DATA).
ANALYSIS SHOWS LARGE POSITIVE f1ARGill.
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AlIALYSES ADEQUATELY ACCOUNTS FOR FRETilNG.
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,1g 3 l-GUIDE TUBE WEAR 1
DESIGN BASES l
MAINTAIN FUEL ASSEMBLY STRUCTURAL INTEGRITY AND CONTROL R0D INSERTABILITY, DESIGN EVALUATION
_SEE BAW 1623
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OBSERVED WEAR IS SMALL AND WELL WITHIN ALLOWABLE' WEAR.
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FIC 9 ALLOWABLE WEAR WEAR TYPE LOAD CONDITION ALLOWABLE WEAR
- UNIFORM
- HANDLING 60%
ONE SIDED HANDLING 100%-
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- FUEL ASSEMBLY AVERAGE Ill PERCEll, OF WALL THICKNESS 1
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1 W e NUMBEROFGlilDETUBES,EEASUREDWAS185
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MEASUREMEllTS FROM HIGH, INTERMEDIATE AND 7-LOW FLOW Pl. ANTS
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MAXIMUM IllDIVIDUAL GUIDE TUBE WEAR ilAS 57%
MAXIMUM FUEL ASSEMBLY AVERAGE GUIDE TUBE WEAR WAS 111.9%
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EXPEC'3] B:HAV::03 OXIDE LAYER APPROXIMITELY 1 MIL. AT' PROJECTED BURNUP (37,000 MWD /MTU)
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NO ACCELERATION OF OXIDATION i
THRU PROJECTED BURNUP 4
i CRUD THICKNESS IS APPROXIMATELY 1 MIL DURING THE FIRST REACTOR CYCLE t
CRUD THICKNESS DIMENISHES DURING SUBSEQUENT CYCLES 9
CRUD DEPOSITION RATE IS VERY LOW ON
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FRESH FUEL DURING SUBSEQUENT CYCLES
FIC 12 OXIDE LAYER THICKNESS OF B&W FUEL RODS AT
_,VARIOUS AXIAL LOCATIONS
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(1) KWU OXIDE L AYER BOUNDS FROM:
i EPRI NP-1472 " REVIEW 0F PWR FUEL ROD WATERSIDE C_0RROS10N
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BEHAVIOR", AUGUST 1980
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o.OCONEE 1, CORE 1 FUEL
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e COMPARISON OF CRUD FLAKE THICKNESS e
CYCLES TbbkS' f!S l
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COMPARIS0N OF CRUD SAMPLE ACTIVITIES IS0 TOPS ACTIVITY,, k: 1 CR-51 MN-54 C0-58 C0-60 ZR-95
-1sr CYCLE 150 150 2100 355 16 2ND CYCLE-380
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TREATMENT OF O.XI DATI ON/ CRUD IN -THERMAL ANALYSIS.
l Benchmarking (TACO, TACO-2) assumes no crud or oxidation ~
Benchmarking data is conservatively predicted theref ore crud and oxidation are not explicitly included in thermal analyses TACO is more conservative than TACO-2 f or limiting condition TACO-2 conservatism has been quantified in recent NRC discussions
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FIG 16 OXIDATION / CRUD BUILDUP Design Basis Oxidation / Crud buildup sliould'-
not cause 'an excessive increase in temperature -
Design Limit No explicit limit value is specified - eff ects of crud and oxide buildup are included implicitly in f ucl therrnal perf ormance evaluations 8
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Design Basis -
Cladding should I
not fail due to excessive internal pressure Design Criteria - Internal Pressure shouldibe -less than noininal. system r
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pressure during normal-operation Design Evaluation - Generic analysis shows that int ernal pressure is less than system pressure criteria is met O
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F nc 19 FISSION GAS RELEASE i
AT HIGH BURNUP FUEL THERMAL PERFORMANCE EVALUATION TACU code ~ = BIW--10087A
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Ref erenced in Midland FSAR and includes t~emperature--dependent FGR model with burnup'-dependent NRC mode 1 (4.2.1.3.1)~
TACO-2 code - BAW-10141P Under review, includes new temperature
-burnup dependdnt Fission Gas Release Model e
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FUEL' ROD INTERNAL
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PRESSURE ANALYSIS Discussion provided in Midland FSAR Sections 4.2.1.3.1 and 4.2.3.i.2
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- Presstire less than System Pressure 9
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burnup of f uel. batch containing limiting (maximum power) assembly
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Example - M-idland Core 1 Maximum Maximum Batch' - Bdrdnfi ~ ~ FAH.
I
~
1 16720 1.39 2
177G0 1.58 1.66. E L c..
/..'V 3
14650 e
PENALTY Current Method 0.8% less 1.0F, = 0%.
Proposed Method O.1% 1ess 1.0% = 0%
9
~.s M
-_- _- A*- i : : __t * '- - -
' - ~ ~
nc 22 FUEL ROD BOW DNBR PENALTY V
Burnup Current Proposed
(%)
(%)
O.
.00.-
..00 5000
.05 -
.00.
10000
.35 :.
.00..
15000
.75
.05 20000 1.50
.25 25000 2.35
.65 '
30600 3.40 1.40
~
35000 4.65 2.30 40000 6.00 3.30 GENERIC OFFSET APPLIED 1%
-Pitch Reduction Factor (1) Response to Question 221.14 (2) BAW-10147P 4
e b
d at M4k.
M wrsa % 4 ) e a empG h et t '
- "***T'Cax
.* 6
no n l
FUEL ROD BOW STATUS Current DNBR. penalty e.valuatio.ns
-are based.on interim mcidel accepted
'~
by NRC' 'in letter -
~
.:t L S Rubens ~tein to J H Taylor October 18, 1979 Proposed DNBR penalty evaluation is provided..in. Topical Report BAW-10147P (currently under review by NRC)
FIG 24 3
FUEL ROD BOW DNBR_ PENALTY..T.
Application
~
Penalty 1s applied to DNBR limit value for evaluation of Condition I and II events and f or determination of Reactor
^
~
Protection System Limits Midland Core 1 -
DNBR Limit Value 1.30 BAW-2 ^
Rod Bow DNBR Pena'lty 0%
Revised DNBR Limit 1.30 (1.30 x 1.0) 4 S
a
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<. :.. L..* $~ -.
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..s. J S101.ZR GA? AXAISTS
. DESIGN BASIS-NO AXIAL INTERFERENCE.
~
A.
FUEL R0D TO ENDFITTING 1.0%
~
B.
FUEL ROD TO REACTOR INTERNALS 0.6%
DESIGN EVALUATION-NO INTERFERENCE FOR PROJECTED BURNUPS i
\\
f, FIG 26 NUREG-0630 IMPLEMENTATION A)
REVISED RUPTURE, STRAIN'AliD BLOCKAGE CORRELATIONS IN CRAFT 2, BAW-10092 AND THETA 1-6, BAU-10094 FOR SLOW AND FAST RAMP RATES.
-B). ANINTERPOLATIONALGORITH.".flILLBEDEVELOPEDINTHEAB0VECODES L
FOR RAMPS BETWEEll SLOW AND. FAST..
J C)
AU ALGORITHM TO CALCULATE EFFECTIVE RAMP. RATES-WILL BE DEVELOPED Ill THE AB0VE.. CODESg D) 'SENSITIVITYSTUDIES1 HILL'BEPERFORMeDTODETERijlNE'IfyCTONCODE
~ ~ " '
~
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RESULTS.
~
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E)
SELECTED TRADEOFFS WILL BE INCLUDED TO 0FFSET IMPACT.
C-R A F i 2.
1~." 0 M I* 5 F)
REVISIONS WILL BE FILED WITH THE'NRC STAFF FOR BAW-10092, 3AW-10094 AND THE. EVALUATION MODEL, BAW-10104, e : u.
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EVALUATION OF FUEL ASSElBLY
. STRUCTURAL, RESPONSE-TO EXTERNALLY' APPLIED LOADS a
-1.
FSAR REFLECTS BAW 10035 2.
N5NANLYSiS'ADiiREssASYMMETRICELOCA, COMBINED l
SSE AND LOCAEAND.MIDl.AND UNIQUE CHARACTERISTICS SRP 14.2EAPPENDIX A (NUREG-0800)
BAW 10133P~(REV. 1)
~
EXPECT POSITIVE MARGINS t
G e
n i
...ai..-_
^
~
FIG 29 LICENSING SUBMITTALS
(
C REVIEW ~
- W BAW10133P(REV.1)(NT 1.
RESPONSES TO QUESTIONS ~
RESULTS OF SAMPLE PROBLEM RESULTS OF SENSITIVITY STUDY 99?
COMBINATION OF LOADS k
GRID STREllGTH 2.
BAW 1621 RESPONSES TO QUESTIONS p
w ~+ -
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EX:?3CTE] 33SCTS
~~
SE::SY::C LOCA ANALYS::S
- 1. ANALYSIS IN BAW 10133P(REV 1)
IS FOR SSE=0.3g AND ASYMMETRIC LOCA
- 2. MIDLAND. SSE=0.12g
~
- 4. NEW MIDLAND ANALYSIS EXPECTED TO SHOW FUEL ASEMBLY TO HAVE SIGNIFICANT POSITIVE MARGIN 4
4
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MEETING
SUMMARY
DISTRIBUTION
- Docket File - so - 3 2 cf G. Lear NRC/PDR S. Pawlicki
' Local PDR V. Benaroya TIC /NSIC/ TERA Z. Rosztoczy LB #4 r/f W. Haass H. Denton D. Muller NOV g geg t E. Case R. Ballard D. Eisenhut W. Regan R. Purple R. Mattson B. J. Youngblood P. Check A. Schwencer
- 0. Parr F. Miraglia F. Rosa J. Miller W. Butler G. Lainas W. Kreger R. Vollmer R. Houston J. P. Knight W. Gammill R. Bosnak L. Rubenstein F. Schauer T. Speis R. E. Jackson W. Johnston SE Bi' S: "Our A
/Q:[n-'
ACRS (16)
F. Schroeder
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% p.
/e>'I' JL-R. Tedesco.
D. Skovholt o
.J N. Hughes M. Ernst.
K. Kniel I
NOV101981* E NRC
Participants:
G. Knighton 12
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A. Thadani ccenscei
-p D. Hood D. Tondi I-Qj R. O. Meyer J. Kramer
'MS ly a v \\,8',
D. Powers D. Vassallo M. Takar P. Collins R. Gupta D. Ziemann W. Jensen F. Congel R. Voglewede J. Stolz M. Srinivasan R. Baer
.bcc: Applicant & Service List C. Berlinger x
E. Adensam Project Manager o _ ynnci Licensing Assistant M. Duncan 9
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