ML20211M527
| ML20211M527 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/30/1986 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML19292G465 | List: |
| References | |
| CEN-347(O)-NP, TAC-63967, NUDOCS 8612170328 | |
| Download: ML20211M527 (62) | |
Text
9 CEN-347(0)-NP
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Omaha Batch M Reload Fuel Design Report 1
l November 1986 l
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'J Combustion Engineering, Inc.
Nuclear Power Systems Windsor, Connecticut A
oOk285 PDR
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LEGAL NOTICE l , ,
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THIE REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED '
SY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING '
NOR ANY PERSON ACTING ON ITS BEHALF:
A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR llWWED INCLUDING YHE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTA81UTY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR
- 8. ASSUMES ANY LIA81UTIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THl3 REPORT.
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Table of Contents Section 1.0 Introduction and Summary 1 1
1.1 Introduction 1
1.2 Summary 4
Fuel System Desigs Description 2 2.0 2.1 Fuel Assembly Description 2 2.2 Fuel Rod Description 3 2.3 Poisoe Rod Description 3 Fuel System Design Bases 3 3.0 3.1 Fuel Assembly Design Bases 3 Design Conditions 4 3.1.1 3.1.2 Fuel Assembly Structural Integrity criteria 5
7 3.1.3 Naterial Selection Control Element Assembly Guide Tubes 7 3.1.4 Zircaloy-4 Rar Stock 8 3.1.5 3.1.6 Zircaloy-4 Strip Stock 8 9
3.1.7 St=4n1=es Steel Castings 3.1.8 Stainless Steel Tubing and Bar Stock 9 3.1.9 Inconal I-750 compression Springs 9 3.1.10 Inconal 625 Spacer Grid Scrip Material 9 3.2 Fuel Rod Design Basas 10 3.2.1 Fuel Rod Design Limits 10 Fuel Rod Cladding Yroperties 12 3.2.2 Fuel Rod Cladding Dimensional Requirements 13 3.2.3 3.2.4 Feti Rod Cladding Metallurgical Properties 13 3.2.5 Fuel Rod Cladding Chemical Properties 14 Fuel Rod Component Properties 14 3.2.6 3.2.6.1 Zirealoy-4 Bar Stock 14 3.2.6.2 Stainless Steel Compression Springs 14 3.2,7 UO Fuel Pellet Properties 14 2
3.2.7.1 Chemical Composition 15 15 3.2.7.2 Microstructure 3.2.7.3 Density 16 3.2.7.4 Thernal Properties 16 3.2.7.5 Mechanical Properties 17 Fuel Rod Pressurization 17 3.2.8 3.2.9 Design Linit on Fuel Melting 18 1
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l Table of Contents (Cont'd.)
Section 3.3 Burnable Poison Rod Design Bases 18 3.3.1 Burnable Poison Rod Design Limits 18 3.3.2 Burnable Poison Rod Cladding Properties 19 3.3.3 A1 0 -B C Burnable Poison ?ellet Properties 23 4 19 3.3.3.1 Thermal-Physical Properties 19 3.3.3.2 Irradiation Properties 21 3.3.3.3 Chemical Prcperties 22 4.0 Fuel System Design Evaluation 22 4.1 Tual Cladding Collapse 23 4.2 Irradiation Induced Dimensions 1 Changes (Growth) 24 4.3 Cladding Strain and Tatigua Analysis 24 4.4 Maximum Fuel Rod Internal Pressure 25 4.4.1 Design Methods 25 4.4.2 Results 25 4.5 Clad Boving 26 4.6 Corrosion and Hydrogen Pickup 27 4.7 Pellet-Clad Interaction 27 4.3 Power to Fuel Centerline Malt 27 5.0 Fuel Bundle Assembly Compatability 27 5.1 Hydraulic Compatability 27 5.2 Mechanical Compacability 28 References 29 Tables 30-44 Figures 45-58 i .
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1.0 INTRODUCTION
AND
SUMMARY
1.1 Introdtsction The purpose of this document is to provide a suunnary of the design,
. design bases and design evaluations of the Batch M reload fuel bundle assemblies for Omaha. The Batch M fuel bundle assemblies are designed to be mechanically, thermally, and hydraulically compatible with the Exxon fuel bundle assemblies remaining in the core. The
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Batch M reload fuel assemblies also comply with both past and current nuclear regulatory requirements.
. This document will b2 presented in four sections:
Section 2.0 describes the fuel system design Section 3.0 describes the fuel system design bases Section 4.0 describes the fuel system design evaluation performed Section 5.0 describes the compatability between C-E and Exxon fuel bundles 1.2 Summary The Batch M fuel design is similar to the previous C-E Batch G fuel design with a few significant differences. For Batch M the cold beginning-of-life shoulder gap (space between the tops of the fuel rod end caps and the underside of the upper fitting flow plate) has beenincreasedfros[ peh toc ] inch. To enable the increase in shoulder gap ofC ]inchtheoveralllengthofthe fuel rod has been decreased byC, Jinch and the height of the lower and fitting has been decreased by [ Jinch.
The Batch M fuel design when compared to the Batch G design has eight poison rods within the grid pattern in place of the fuel rods.
The overall length of the poison rod assembly isC Jinch shorter which provides beginning-of-life shoulder gap ofC, ] inch.
,The lower and fitting design has also been modified by includine _.
(See Figure 1)
Becauseofthe( )inchreductioninthelowerendfittings -
height, the design of the lower end fitting to guide tube connection also was changed. The lower and fitting to gechG,thelastC-Ereload,consistedof(guidetubeconnection3 See Figure 2 which shows a comparison of this concept with~
,he c M concept.
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4 The Batch M lower and fitting to guide tube connection consists of ,
(see Figure 2).
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- In addition to the lower and fitting changea described, the Batch M design has slots =neh h ad on the bottom of the outer post legs as
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compared to windows ==-h " -d in the post less in Batch G. This change simplifies the nachining operation while still allowing access to weld the pins in the lower and fitting.
For infotsation. Table 4 lists the design drawings used in the fabrication of both the C-E Batch G and Batch M fuel bundle assemblies. Accompanying the drsvings listing is a brief description, when applicable, of any design changes between the two batches.
2.0 FtTII, ST5 TEN DESIGN DESCRIPTICN 2.1 Fuel Assembiv Descriocion The Batch M reload fuel assembly has a 14x14 square array with 176
' rods and 5 guide tubes. The guide tubes, spacer grids and upper and lower and fittings form the structural frame of the assembly.
The upper and fitting is an assembly consisting of evo cast 304 stainless -
steel plates, five machined 304 stainless steel posts and ona helical Inconal I-750 spring nested around the center post (see Figure 4). The upper and fitting assembly serves the following functions: (1) it is an attachment for the guide tubes. (2) it is the fuel assembly lif ting fixture and (3) it is the alignment device l
l that locates the upper portion of the fuel assembly in the reactor core. The lower cast place locates the top ands of the guide cubes and is designed to retain the fuel rods in the event of certain l
postulated accident conditions. The lower and fitting (See Figure
- 1) is a single piece 304 stainless steel casting that has the
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following features:
- 1) a plate with[]3, 2)circular flew legs four support holes with[
pins added to the legs which serve as both seating and alignment aid in aligning and seatin g devicesand3)[
the fuel assemblies in the reactor.
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The eight Zircaloy-4 spacer grid assemblies and five guide tubes are joined by velding at axial locations established to meet apprargriate design criteria for structural integrity and flow distribution. The .
locations also are established to ensure that the Batch M fuel bundle assenh11es are meewa="y and hydraulically compatible vi h the reload fuel bundlas they are loaded with to maka up a full core complement (see Figure 4).
. 2.2 Fuel Red Description The fuel rods consist of slightly enriched UO., cylindrical carsmic pellets, a round wire Type 302 stainless steal compression spring, and two alumina spacer discs located at each and of the fuel column.
All of these components are within a Zircaloy-4 tube seal welded with Zircaloy-4 and caps (See Figure 5). The fuel rod are isternany pressurized with helium during assembly to ]SIG.
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The UO d h ffects of therme$expansion penets areanddished at both ends to accommo ate t e eThe penet density is fuel swelling.
oracica1 density resulting in a column stack density of approximately 3
,]sm/cm The compression spring located at the top of the fuel pellet column maintains the column in its proper positiou during handling and shipping.
The alumina spacer disc at the lower and of the fuel rod reduces the lower and esp temperature, while the upper spacer dise prevents UO chips. if present. from entering the pianum region. Thefuelrodhienum.
which is located above the penet column. provides space for axial l
thermal differential expansion of the fuel column and accomodates the initial helium loading and evolved fission gases.
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l 2.3 Poison Red Descrintion The poison rods are structur2My identical to the fuel rods except they contain poison penets in place of UO2 Pellets and the column of burnable poison penets is slightly shorter than the column of fuel pellets (114.000 inches of poison penet length vs. 128.000 inches of fuel penet length) . The poison material for Batch M is almnin with uniformly dispered baron carbide particles. The balsace of the column consists of C ]locatedatthetopand bottom ends of the poison column (see Figure 6) . The poison rods are internaMy pressurized with helium during assembly to( )PSIG.
3.0 FUEL STSS DESIGN 3ASES
. 3.1 Fuel Assembit Design Bases The fuel asaemblies are required to meet design criteria for each design condition listed below to assure that the functional requirements are met. Except where specifically noted, the design bases presented in this section are consistant with those used for previous designs. .
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3.1.1 Desisn conditions 3.1.1.1 Nonoveration sud Normal Oseration (Condition I) ,
Condition I situations are those which are planned or expected to occur in the course of handling intial shipping, storage reactor servicing and power operation (including maneuvering of the plant). i Condition I situations must be accomodated without fuel assembly failure and without any affect which would lead to a restriction on l subsequent operation of the fuel assembly. The guidelines stated i I
below are used to decarmina loads during Condition I situations: 1
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A. Handling and Fresh Fuel Shineing l Loads correspond to the ==w%= possible axial and lateral loads and accelerations imposed on the fuel assembly by shipping and M14ag equipment during these periods, assuming that there is no abnormal contact between the fuel assembly and any surface, nor any equipment malfunction. Irradiation effects on material properties are considered when analyzing the effects of h=.nd14ng loads which occur during refueling.
- h. Storate Loads assumed for both new and irradiated fuel assemblias must reflect storage conditions of temperature, chemistry, means of support and duration of storage.
C. Reactor Servicing Loads assumed for the fuel assembly must reflect those encountered during refueling and reconstitution.
D. Power Ooeration Loads for these conditions are derived from operating states encountered during transient and steady-state operation in the design power range. (Hot operational testing, system startup.
hot standby, operator-controlled transients within specified rate limits and system shutdown are included in this category.)
j E. Reactor Tris Loads assumed correspond to those produced in the fuel assembly by control alament assembly (CIA) motion and deceleration.
3.1.1.2 Upset Condition (condition II)
Condition II situations are unplanned event's including operating
- basis earthquake (CBE) which may occur with moderate frequency during the life of the plant. The fuel assembly design should have the capability to withstand any upset condition with margin to mechanical failure and with no permanent effects which would prevent continued normal operation.
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3.1.1.3 Imerzenev Conditions (Condition III) .
Condition III events are unplanned incidents including miner fuel handling accidents which might occur infrequently during plant life.
2nd M -= 4 - =1 failure must be prevented fer any Condition III event in any area not subject to extreme local conditions (e.g., in any rod not imediately adjacent to the impact surface during fuel u d'4=: accident).
3.1.1.4 Faulted Conditions (Condition TV)
Condition IV incidents are postulated events plus the Safa Shutdown Earthquake (SSI) . LOCA (Mechanical Excitation only), combined SSI and LOCA, and major fuel handling accident vboss consequences are such that integrity and operability of the nuclear energy system may be impaired. Mechanical fuel failures are permitted, but they must not impair the operation of the engineered safety features (EST) systems to mitigate the consequences of the postulated event.
3.1.2 Fuel Assembiv Structural Integrity Criteria For each of the design conditions, there are criteria which apply to the fuel assembly and components with the exception of fuel rods.
These criteria are listed below and give the allowable stresses and functions 1 requirements for each design condition. Criteria for fuel rods are discussed separately.
3.1.2.1 Desien conditions I and II P, j 5, F, + Pb d F,S, Under cyclic loading conditions, stresses nust be such that the cumulative fatigue damage factor does not exceed 0.8. The cumulative damage factor is defined as the sum of the ratios of the l
number of cycles at a given cyclic stress (or strain) condition to the maxinena number of cycles permitted for that condition. The factor of 0.8 is chosen to provide margin in the design.
3.1.2.2 Desian Condition III P,j 1.5 5 ,
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. P, + Pb d 1.5 F,$ ,
3.1.2.3 Design Condition IV s
P <S' m- a
? +P < TsaS' a b where S' = smaller value of 2.4 5 ,or 0.7 S,.
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(1) if the equivalent diameter pipe break in the LOCA does not exceed 0.5 square foot, the fuel assembly defe: ration shall be ,
limited to a value not exceeding the deformation which would preclude satisfactory insertion of the CIAs.
(2) Tor pipe break sizes greater than 0.5 square foot, deformation of structural components is limited to maintain the fuel in a coolable array. CIA insertion is not required for these events as the appropriate safety analyses do not take credit for CIA insertion.
~ (3) For the upper and fitring springs, calculated shear stress must not exceed the =N== yield strass in shear.
3.1.2.4 Nomenclature The symbols used in defining the allowable stress levels are as follows:
P ,= Calculated general primary membrane stress (*
P = Calculated primary bending stress b
5, = Design stress intensity value asfefined by Section III ASE Boiler and Pressure Vessel Code' S , = Minimum unirradiated ultimate tensile strength I
T, = Shape factor cogsponding to the particular cross section l being analyzed
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l S' = Design stress intensity value for faulted conditions The definition of S' as the lesser value of 2.4 S and 0.7 5 is contained in the ASEE Boiler and Pressure Vessel bde.Section III.
(a) P and P are defined by Section III. ASE Boiler and Pressure b
Vessel Code.
(b) With the exception of zirconium base alloys, the design stress intensity values, S d by the Code are determined in t3e, sameof mannar materials not as the tabulate Code. The design ,
stress intensity of zirconium base alloys shall not exceed two-thirds of the unirradiated minimum yield strength at temperature. Basing the design stress intensity on the unirradiated yield strength is conservative because the yield strength of zirealoy increases with irradiation. The use of the two thirds factor ensures 50I margin to component yielding in response to primary stresses. This 50I margin together with its application to the minimum unirradiated properties and the general conservatism applied in tLi establishment of design conditions is sufficient to ensure an adequate design.
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(c) The shape factor, T , is defined as the ratio of the " plastic" acaent (all fibers just at the yield stress) to the initia.
yield amount (extreme fiber at the yield stress and an other fibers stressed in proportion to chair distance from the neutral axis).
3.1.3 Material Selection
- The fuel assembly grid case structure consists of 8 Ziresloy-4 spacer grids, 1 Inconel 625 spacer grid (at the lower and), 5 Zirtaloy-4 guide tubes, 2 stainless steel and fittings, and 1 Inconal I-750 coil spring. Zirealoy-4, selected for fuel and poison rod cladding, poison rod spacer tubes, guide tubes and spacer grids, has a low neutron absorption cross section, and high corrosion resistance to reactor water environment. Also there is little
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reactics between the cladding and fuel or fission products. l
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The bottom spacer grid is of Inconal 625 and is velded to the lower and fitting. In this region of local inlet turbulance, Inconal 625 j was selected rather than Zircaloy-4 to provide additional strength and relaxation resistance. Inconal 625 is a very strong nacerial with good ductility, corrosion resistance and stability under irradiation at temperatures below 1000F.
The fuel assembly upper and lower and fitting are of cast 304 stainless steel and the upper and fitting posts are 304 stainless steel anchtned components. This material was selected based on considerations of adequata strength and high-corrosion resistanca.
Also 304 stainless steel has been used successfully in almost an pressurized veter reactor environments, including au currently operating C-E reactors.
3.1.4 Control Elemene Assembiv Guide Tubes All CIA guide tubes are manuf actured in accordance with ASIM 3353, Wrought Zirconium and Zirconium Anoy Seamless and Weldad Tubes for Nuclear Servica, with the following exceptions and/or additions:
3.1.4.1 Chemical Procerties Additional limits are placed on oxygen.
3.1.4.2 Mechanical Prooerties A. Flare A section of tube, between 2 and 4 inches in length, shall be flared with a tool having a 60-degree included angle until the outside diameter has increased by 15Z. The flared tube shan show no cracking when amined with the unaided eye.
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l 3.1.4.3 Dimensional Recuirements Permissible Tolerance -
Dimension (in.)
CD 20.003 ID 20.005 3.1.5 Zireslov-4 Bar Stock All Zircaloy-4 bar stock is fabricated in accordance with Grade
- RA-2, ASD1 B351, Hot-Rolled and Cold-Finished Zirconium and '
Zirconium Alloy Bars, Rod and Wire for Nuclear Application, with the fallowing exceptions and/or additions:
3.1.5.1 Chemical Procerties Additional limits are placed on oxygen and silicon content.
3.1.5.2 Metallurgical Proeerties A. Crain Size The max 4==== average grain size is restricted.
3.1.6 Zircalow-4 Strip Stock All Zircaloy-4 strip stock is fabricated in accordance with Grada RA-2, ASTM B352, Zirconium aan Zirconium Alloy Sheet, Strip and Place for Nuclear Application, with the following exceptions and/or additions:
3.1.6.1 Chemical Properties Additional limits are placed on oxygen and silicon content.
3.1.6.2 Metallurtical Procerties A. Crain Size The ==rimum average grain size is restricted.
3.1.6.3 Mechanical Pronerties A. Bend Each sample shall be bent using a three point type, guided bend test fixture similar to that described in MAB-192-M Evaluation Test Methods for Refractory Metal Sheet Material Paragraph 5.2.2, published by Division of Engineering and Industrial Research, National Academy of Sciences National Research 8
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. Council April 22, 1963. The semple shan be bent 180* using a hand anvil with a radius equal to twice the sheet thickness. .
After bending, each specisan shan be liquid If dyecracking penetrant occurs l inspected to assure freedom from cracking. !
on any part of the band samples, the coils represented shall be I rejected.
3.1.6.4 Coefficient of The -=1 Ezoansion Axial direction - See Table 2 .
3.1.6.5 Irradiation Procerties The The yield and tensile strengths are enhanced by irradiation.
stress relaxation with irradiation at operating temperatures proceeds at a rapid race until nearly complete.
3.1.7 Stainless Steel Castinss All s*=da1==s steel castings are fabricated in accordance with Grade CF-8, ASTM A744 with the f all
- ring addition:
Chemical Properties Cobalt content is limited.
3.1.8 Stainless Steel Tubing and Bar Stock All stainless steel posts, alignment pins and bolts cre f abricated in accordance M th ASTM A269 and ASTM A276, with the fonoving addition:
Chemical Properties Carbon content is limited on tubing to be velded.
Cobalt content is limited.
3.1.9 Inconel I-750 Comoress un Sering I
All Inconal springs are fabricated in accordance with AMS 5699B.
with the fonoving addition:
Chemical Properties Cobalt content is limited.
Passivation prohibited 3.1.10 Inconal 625 Bottom Soacer Crid Strin Material Inconal spacer grid strip material is procured in accordancealloy with the specification for nickel-chromium molybdenun colunbium plate, sheet, and strip. Specification ASTM B433-72, with the following additional requirements:
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t 3.1.10.1 Chemical Prooerties Cobalt content is limited to 12 ==rd==.
3.1.10.2 soecial Tests A check analysis and a bend test are required.
3.2 yuci Red Design Bases 3.2.1 Fuel Rod Desien Limits
- The fuel cladding is designed to sustain the effects of steady-state and expected transient operating conditions without exceeding acceptable levels of stress and strain. Except where specifically i
noted, the design bases presented in this section are consistant with those used for previous core designs. The fuel rod design accounts for cladding irradiation grovch, external pressure, differential expansion of fuel and clad, fuel swelling, densification, clad creep, fission and other gas releases, initial internal h i4n= pressure, thermal stress, pressure and temperature cycling, and flow-induced vibrations. The structural criteria discussed below are based on the following for the normal, upset, and emergency loading combinations identified in Section 3.1.1.
3.2.1.1 During normal operating and upset conditions, the maxistna primary tensile stress in the Zircaloy clad shall not exceed two-thirds of the minimum unieradiated yield strength of the material at the applicable temperature. The corresponding limit under emergency conditions is the material yield strength.
3.2.1.2 Net unrecoverable circumferencial strain shall not exceed 12 as predicted by computations considering clad creep and fuel-clad interaction effects.
3.2.1.3 The clad vill be initially pressurized with helium to an amount sufficient to prevent gross clad deformation under the combinaci effects of external pressure long-term cresp. The clad design vill not rely on the support of fuel pellets or the holddown spring to prevent gress deformation.
3.2.1.4 Cumulative strain cycling usage, defined as the sina of the ratios of the number of cycles in a given effective strain range ( Ac) to the permitted number (J) at that range vill not exceed 0.8.
i 3.2.1.5 There is no specific limit on lateral fuel rod deflection for
- structural integrity considerations except that which is brought about through application of cladding stress criteria. The absence of a specific limit on rod deflection is justified because it is the fuel assembly structure, and not the individual fuel rod. that is the limiting f actor for fuel assembly lateral deflection.
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s 3.2.1.6 Fuel rod internal pre'ssure increases with increasing burnup and
, toward end-of-life the total internal pressure, due to the combined effects of the initial helium fill gas and the released fission gas, can approach values comparable to the external coolant pressure.
The maximum predicted fuel rod internal pressure vin be consistant with the following criteria.
A. The primary stress in the cladding resulting from differential pressure vin not exceed the stress limits specified earlier in this section.
B. The internal pressure vin not cause the cladding to creep outward from the fuel pellet surface while operating at peak linear heat rates up to the fuel design limit during normal operation. In determining compliance with this criterion, internal pressure is calculated for the peak power rod in the reactor, including accounting for the maximum computed fission gas release. In addition, the pellet svening rate (to which the calculated clad creep rate is compared) is based on the observed swelling rate of " restrained" penets (i.e., penets in contact with clad), rather t.han on the greater observed '
svening behavior of penets which are free to expand.
The criteria discussed above do not limit fuel rod internal pressure to values less than the primary coolant pressure and the occurence of positive differential pressures would not adversely affect normal operation if appropriate criteria for
- cladding stress, strain, and strain rate are racisfied.
3.2.1.7 The design limits of the fuel rod cladding, with respect to vibration considerations, are incorporated within the fuel assembly design. It is a requirement that the spacer grid intervals, in conjunction with the fuel rod stiffness, be such that fuel rod 1 vibration, as a result of mechanical or flow induced excitation, does not result in excessive wear of the fuel rod cladding at the spacer grid contact areas.
3.2.1.8 During normal operating and upset conditions (Design Conditions I and II) oxidation and crud build-up have not been observed as a problem as long as C-E coolant chemistry limits are maintained.
Therefore, no specific criteria have been defined for oxidation and crud build-up.
3.2.1.9 Fuel rod burnup experience has proven that any specific criterion on allowable deflections (boving), with respect to the effects which such deflections might have on thermal hydraulic performance, is not necessary beyond the initial fuel rod positioning requirements required of the grids. This variation in spacing is accounted.for
. in thermal-hydraulic analysis through the introduction of hot channel factors in calculating the max 1=um enthalpy rise in calculating DNBR. This adjustment is called the pitch, bowing, and clad diameter enthalpy rise factor, which is conservatively applied
!. to simulate a reduced flow area along the entire channel length.
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3.2.1.10 For axial growth considerations the fuel assemoly is desde ed on the bases of maintaining: 1) an adequata shoulder gap between the fuel (and poison) rods and the upper end fitring, and 2) sufficiant clearance between the fuel assembly and the upper guide structure throughout the expected life (burnup) of the fuel assembly. Spacar grida are designed to allow axial rod growth without introducing ,
significant restraints that would potentially enhanca fuel rod bowing.
There is no critarion for axial growth per se; however, adequate clearancas are maintained on fuel assembly structural components.
- fuel and poison rods, control elemant assemblies, neutron sources, I and in-core instruments to ensure functionability for their respective lifseines. Adequate cisarance is evaluated on the basis that under adverse design conditions a clearance of greater than zero shall be maintained.
3.2.1.11 The design limits of the fuel rod and plenum spring are as follows:
(1) Shaar strain in the planum spring must not exceed 90% of
=4=4 - = ultimate strain in shear.
(2) Deflection of the plenum spring must not cause it to reach solid height or to seize within the clad I.D. at any cias during the fuel design lifetime.
(3) Pellet column axial action relative to the clad must be prevented for axial accelerations of the rod up to 5 g's during initial shipping and hm*d"ng.
3.2.2 Fuel Rod Claddine Mechanical Pronerties 3.2.2.1 Modulus of C.astietty Young's Modulus is given in Table 2.
3.2.2.2 Poisson's Ratio Poisson's Ratio is given in Table 2.
3.2.2.3 Thermal Coefficient in Exnansion Thermal Coefficient of Expansion in the diametral and axial directions is given in Tabla 2.
3.2.2.4 Yield Strength Yield strength is given in Tabla 2.
The cladding stress limits identified in Section 3.2.1.1 are based on values taken from the etMaum yield strength curve at the appropriata temperatures. The limits are applied over the entire fuel lifetime, during conditions of reactor heatup and cooldown, steady state operation, and normal power cycling.
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s 3.2.2.5 Citimate Strength Ultimate tensile strength is given in Tabla 2. -
3.2.2.6 Uniforn Tensile Strain Uniform tensile strain is given in Table 2. f I
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3.2.2.7 Flare A section of tube, approximata 2 to 4 inches in length, shall be
- flared with a tool have a 60-degree included angle until the outside diameter has increased by 152. The flared tube shall show no cracking when ====4ad with the unaided eye.
3.2.2.8 Hvdrostatic Burst Test The cladding specification requires that two suspies from sach lac ,
of cladding be subjected to room temperature hydrostatic burst tests. To be acceptable, the burst pressure must exceed a minimum value, based on the cladding geometry and specified tensile properties, and the circumferential elongation must exceed a prescribed =4n4=== valua of 127..
3.2.3 Fuel Rod Cladding Dimensional Eeouirements (1) Tube straightness is limited to 0.010 in./f t, and inside f
diameter and vall thickness are tightly controlled.
(2) Ovality is measured as the difference between ==rd== and
= 4 n 4 == inside diameters and is acceptable if within the diameter tolerances. .
(3) Outside diameter is specified as 0.440 inches.
(4) Inside diameter is specified as 0.38a inches.
(5) Eccentricity is defined as the diffarance between maximum and
~4"4-~= vall thickness at a cross-section, and is specified as 0.004 inches ==rd==. ,
(6) Wall thickness is specified as 0.026 inches =4"4== (the nominal value reported aisewhere is based on the acainal CD and ID). -
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(7) The overall rod langth tolerance is specified as linches.
. .8 3.2.4 Fuel Rod Claddine !iec211urriesl P enerties 13
4 3.2.4.1 Hvdride Orientation I
A restriczion is placed on the hydride orientation factor for any third of the tube cross-section (inside, middle, or outside) . The hydride orientation factor, as defined as the ratio of the number of ,
radially oriented hydride platelets to the total number of hydride platelets, shall not exceed 0.3. The independent evaluation of three portions of the cross-section is included to allow for the possiblity that hydride orientation any not be uniform across the entire cross-section.
3.2.5 Fuel Rod Claddine Chemical Properties All fuel rod c1=dA4=! is manufactured in accordance with AS2f 3353, Erought Zirconimum and Zirconium Alloy Seanlass and Welded Tubes for Nuclear Service, except additional limits are placed on oxygen, silicon, and iron content.
3.2.6 Fuel Rod Component Properties Zircalow-4 Bar Stock i
3.2.6.1 All Zircaloy-4 bar stock is fabricated in accordance with ASTM 3351.
Eot-Rolled and Cold-Finished Zirconium and Zirconium Alloy Bars Rod and Wire for Nuclear Application, with the following ezcaption and/or additions:
A. Chemical Properties Additional limits are placed on oxygen and silicon content.
l I 3. Metallurtical Procerties The ==w-= average grain size is restricted.
C. Non Destructive Testing Ultrasonic inspection is required.
3.2.6.2 Sc=4tess Steel Comorassion Serings .
All stainless steel springs are fabricatied in accordance with APJi 5688, Revision F. The dimensions of these springs are:
Fuel Rod Poison Rod Free length 8.237 in. 8.568 in.
0.364 in. 0.366 in.
Outside diameter 0.042 in.
Wire size 0.072 in.
38 36
- Active number of coils 3.50 lb/in.
Spring constant 3.90 lb/in.
UO Fuel Pellet Properties 3.2.7 2 14 l
o 3.2.7.1 Chemical comoosition Salient points regarding the structure, composition, and properties .
of the UO fuel Pellets are discussed in the following subsections.
2 A. Chemical analyses are performed for the followint constituents:
(1) Total Uranium
. (2) Carbon (3) Nitrogen (4) nuorine
~
(5) chlorine (6) Iron (7) Thorium (8) Nickel B. The oxygen-to-uranium ratio is maintained between 2.00 and 2.02. 3 C. The sum of the calcium + aluminum + silicon contents shall not exceed 300 ppa by weight.
D. The sum of the cross-sections of the following impurities shall not exceed a specified equivalent thermal-neutron capture cross-section of natural boren:
(1) Baron (2) Silver (3) Cadmium (4) Gadolinium (5) Europium (6) Samarium (7) Dysprosium E. The total hydrogen content of finished ground pellets is restricted to values specified in manufacturing specifications.
F. The nominal enrichment of the fuel pellets vill be specified and shall be held within 2 0.05 vt% U-235.
3.2.7.2 Microstructurs A. The pellet fabrication process will ==v4=d's the pore content of pellets in a specified range. Acceptable porosity distribution will be determined by con.parison of approved visual standards with photo-eierographs from each pellet lot.
B. The average grain size shall exceed a specified a4 sd-.
15
e 3.2.7.3 Density A. The density of the sintered pellet after grinding shall be between[ 3and[ ]oftheoreticaldensity(D),basedona 3 UO theoretical density of C Jg/cm 2
B. The in-pile stability of the fuel is ensured by the use of an NRC-approved out-of-pile test during production.
3.2.7.4 Thermal Procerties A. Thermal Ernansion The thermal expansion of UO7 is described by the following temperature dependent equations:
from 25C up to 2200C
% Linear Expansion = (-1.723x10~)+
(6.797x10 T) + (2.896x10 ~IT2) above 2200C
~
% Linear Expansion = 0.204 + (3x10 'T) +
~
(2x10 T ) + (10-1073) where T = temperature - degrees Celsius.
B. Thermal Emissivity A value of 0.85 is used for the thermal emissivity of UO2 pellets over the temperature range 800 to 2600 K.
C. Melting Point and Thermal Conductivity l
l The variation of melting point and thermal conductivity with burnup is discussed in CENPD-139-P.
Spec c sat f 0 2 D.
The specific heat of UO2 is described by the following temperature dependent equations:
- T < 2240F 6
~3 T-3.2432 x 10 C = 49.67 + 2.2784 I 10 P
(t + 460 )2 16
4 T'J,2240F 2
C = -126.07 + 0.2621T - 1.399 I 15'T
-1274
+ 3.1786 x 10'8T 3 - 2.483 x 10 Where:
C = specific heat 3TU/ft - *F T = camparature. *7 3.2.7.5 Mechanical Procerties A. Yount's Modulus of Elasticity The static modulus of elasticity of unirradiayed fuel of 97* TD and deformed under a strain rate of 0.097 hr is given by:
6 E = 14.22 (1.6715 x 10 - 924.4T) where; E = sodulus of elasticity in pai.
T = temperature in *C in the range of 1000 to 1700*C.
B. Poisson's Ratio The Poisson's Ratio of polycrysnlidne UO 2 has value of 0.32 at 25'C. Assuming the decrease is linear, the temperature dependence of the Poisson's Ratio is given by:
~
v = 0.32 - 1.8 x 10 (T-25) where;-
v = Poisson's Ratio T = temperature in *C in the range of 25 to 1800*C At temperature above 1800*C, a constant value of 0.29 is used for Poisson's Racio.
C. Yield Stress - (Not Applicable)
D. Ultimate Stress - (Not Applicable)
E. Uniform Ultimate Strain - (Not Applicable)
. 3.2.8 Fuel Red Pressuri:stion Fuel rods are initially pressurized with helium for two reasons:
17
(1) Preclude clad collapse during the design life of the fuel. The internal pressurization, by reducing stresses from differential pressure, extends the time required to produce creep collapse beyond the required service life of the fuel.
(2) Improve the thermal conductivity of the pellet-to-clad gap within the fuel rod. Helium has higher coefficient of thermal conductivity than the gaseous fission products.
The initial helium fill pressure vill be ( ) SIG. This initial fill pressure vill be sufficient to prevent clad collapse, and vill produce a maximum EOL internal pressure consistent with the criteria.
3.2.9 Desien Limit on Fuel Meltine The peak temperature of the fuel shall be less than the melting point during steady state-operation and ,1uring occurances of moderate frequency (Condition I and II).
3.3 Burnable Poison Rod Design Bases i
The burnable poison rod design accounts for. external pressure, diff erential expansion of pellets and clad, pellet swelling, clad creep,
! helium gas release, initial internal helium pressure, thermal stress, and flow-induced vibrations. Except where specifically noted, the design bases presented in this section are consistant with those used for j
previous CI designs.
3.3.1 Burnable Poison Rod Desien Limits l The structural criteria for the normal, upset and emergency loading l
combinations are as follows:
(1) During normal operating and upset conditions, the maximum primary tensile stress in_the Zircaloy clad shall not exceed two-thirds of the minimum unirradiated yield strength of the material at the applicable temperature. The corresponding limit under emergency conditions is the material yield strength.
(2) Net unrecoverable circumferential strain shall not exceed 12 as predicted by computations considering clad creep and poison pellet swelling effects, t
l (3) The clad vill be initially pressurized with helium to an amount sufficient to prevent gross clad deformation under the combined effects of external pressure and long-term creep. The clad design will not rely on the support of pellets or the holddown spring to l
- yrevent gross deformation. The initial fill pressure vill be L, ]PSIG.
l (4) The design limits of the poison rod end plenum spring are as l
follows:
18 l
.(i) Shear strain in the plenum spring must not exceed 90% of min 4=n= ultimate strain in shear.
(ii) Deflection of the planum spring must not cause it to reach solid height or to seize within the clad I.D. at any time during the fuel design lifetime.
(iii) Pellet column axial motion relative to the clad must be prevented for axial accelerations of the rod up to 1 g during initial shipping and handling.
~
3.3.2 Burnable Poison Rod Cladding Procerties Cladding tubes for burnable poison rods are purchased under the specification for fuel rod cladding tubes. Therefore, the mechanical, metallurgical, chemical, and dimensional properties of the cladding are as discussed in Section 3.2.
A10 -B C Burnable Poison Pellet Properties 3.3.3 23 4 The A1.,0,-Bg C burnable poison pellets used in C-E designed reactors consist 5f a relatively small volume fraction of fine B C particlesTheboronloadkng dispersed in a continuous A1 0 matrix.
byadjustingtheBCconcentha$ionintherangefrom0.7to4.0wt:
(1 to 6.0 v/o). ThebulkdensityoftheA1,0,-B.Cpallacsis specified to be greater than 93% of the calculated theoretical density. Typical pellets have a bulk density of about 95% of theoretical. Many properties of the two-phase Al 20,-B gC mixture, such as thermal expansion, thermal conductivity, an5 specific heat are very similar to the properties of the A1 073 major constitutent.
In contrast, properties such as swelling, heIium release, melting point, and corrosion are dependent on the presence of B i c. The I
operating centerline temperature of burnable poison is Iess than 1150F, with maximum surface temperatures close to 1090F, 3.3.3.1 Thermal-Physical Procerties A. Thermal Expansion The mean thermal expansion coefficients of Al 023 and B C from 0 to1850* Fare 4.9and2.5x10in./in.*y,respectiveky. The thermal expansion of the A1,0,-B C g two-phase mixture can be considered to be essentially the same as the value for the continuous A1.,0, matrix since the dispersed Bgc phase a lower expansion coefficient and occupies only 5 v/o of the available volt.:me . The lov temperature (80 to 250*F) thermal expansion coefficient of A1 0 trradiated at 480, 900, and 1300*F does not change as a 23 result of irradiation. The expansion of a similar material, beryllium oxide, up to 1900*F has also been reported to be relatively unchanged by irradiation. It is therefore appropriate to use the values of thermal expansion measured for A1.,0 for the burnable poison pellets:
3 19
1 I
o 4
Temperature Range Lineer Expansion
('T from 70F to) (%)
400 0.12 600 0.23 800 0.30 1000 0.40 B. Melting Point
- The melting points of Al 023(3710*F) and B C 4 (4400*F) are higher than the melting point of the Zr-4 cladding. No reactions have been reported between the components which would lower the melting point of the pellets to any significant
'- extent. As the B 4C burns up, the lithium atoms formed occupy
, interstitial sites randomly distributed with the B gC isttice, l rather than forming a lithium-rich phase. The solid solution of lithium in B C does not appreciably influence the melting 4
point of the A1 0 -B C pellets, as only a small quantity of lithiumcompounds(Oh5we%)formduringirradiation.
3 It is concluded that the melting point of A1 90 3-Bg C will remain considerably above the' maximum 1150*F 5perating temperature.
C. Thermal Conductivity The thermal conductivity of A1,0 3-B4C was calcualted from the j measured values for A1,03 and B4 C using the Maxwell-Eucken relationship for a continuous matrix phase (A1 023) with spherical dispersed phase (B g C) particles. Because of the high Al 0 e ntent f these mixtures and the similarity in thermal 23 conductivity, the resultant values for Al 0 -B C vere essentially the same as the values for A1,023.34The measured, unirradiated values of thermal conductivity at 750*F are 0.06 cal /s-em *K for B4 C and 0.05 cal /s-cm 'K for Al23 0.
The thermal conductivity of A1 0 after irradiation decreases 23 rapidly as a function of burnup to values of about one-third the unirradiated values. The irradiated values of A1,0 3 -B 4C calculated from the above relationships are given be15w as a function of temperature.
Temperature Thermal Conductivity
(*F) (cal /s-em 'K) 400 0.015 600 0.013 800 0.010 1000 0.008 20
4 D. Specific Heat The specific heat of the A1 70 3-Bg C mixture can be taken to be * ~
essentian y the same as purt A1.,0 3 since the concentration of Bg C is low (6.0 v/o maximum). In addition, the effect of irradiation on specific heat is expected to be small based on experimental evidence from similar materials which do not sustain transmutations as a function of neutron exposure.
The values for Al 0 measured on hadiated samples are ghen below: 23 O
Temperature
(*F) cal /sm 'F 250 0.12 450 0.13 800 0.14 1000 and above 0.15 '
3.3.3.2 Irradiation Properties A. Swaning Al -B C consists of B C particles dispersed in a continuous 4 4 Al matrix, which. occupies more -Jian 95% of the poison pe c. The sweHing of A10 3-BgG depends primarily upon the neutron fluence on thf0 cont.nuous A1.,0 3 matrix and, secondarily, on the B burnup of thE aispersed B _4C phase.
Recent measurements performed on material g gontaining about 2 wt%RG1rrgiatedinaC-EPWRto100%B burnup at a fluence of 2.g4 x 10 nyt (I)0.8 MeV) revealed a diametral swelling of about 1%. Penets similar to the burnable poison used ing-E reactors with up to 3 vt: B C4also sustained about 100% B in burnup. Experimental data on A1.,0 3 re3 gal a diametral swellir.g of about 0.7% at a fluence of 2.2 x 10 nyt (E>0.8 MaV).
Swelling of A1.,0 3 increased linearly g th fluence to 1.8%
diametral aftef In exposure of 6 x 10 nyt (E>0.8MeV).
These data show that At 0 -B C swous somewhat more than M.,0 3 up to a burnup of 40 B C 3(ahout 2 x 10" nyt at E>0.8 MeV)
The C-E design vq' ue of A1.,0 3 g -B C svening rate for fluences less than 2 x 10 while after 2 x 10 ysgreateftHantheswellingrateofA10, fluence the svening rate of Al 23 4 i considered equal to that of Al 0 .
23
~
The data and considerations presented above result in besy-estimate diametral svening values at end-of-lif a (11.4 x 10' nyt. E>0.8 MeV) of about 2% for Al230 and from 2 to 3% for A1 0 -3 C loading.
23 4 21
9 4
- 5. Helium Release Experimental measurer.ents reveal that less than 5% of the helium formed during irradiation vill be released. These measurements were performed on A1.,0,-B g C pellets irradiated at temperatures of 500*F and, and su$s3quently, annealed at 1000*F for 5 days. The helium release in a burnable poison rod which operated for one cycle in a C-E PWR was calcualtad from internal pressure measurements to be less than 5%.
o 3.3.3.3 Chemical Properties A. A123 0 -B'4C Coolant Reaction Should irradiated BgC particles be exposed to reactor coolant, the primary corrosion products that would be formed are horic acid (which is soluble in water), hydrogen, free carbon and a small amount of lithium compounds. The presence of these products in the reactor coolant would not be detrimental to the ,
operation of the plant.
Observations of A1.,0 3 g-B G poison shims have revealed that long term exposure of tE13 material to reactor coolant can result in gradual leaking out of Boron and eventual erroding away of the A1.,0, matrix. However, the race of reaction is such that any reBultant changes in reactivity are very gradual.
B. Chemical Compacabilinv Chemical compatibility between A1.,0 3
-B,C pellets and the burnable poison rod cladding during long-term nornal operation has baan demonstrated by examination of a burnable poison rod from the Maine Yankee reactor. The rod had been exposed to an axial average fluence in excess of 2 x 10'y nyt (>0.821 MeV).
No evidence of a chemical reaction was observed en the cladding I.D.
Short-term chemical compatibility during upset and emergency conditions is demonstrated by the fact that conditions are not favorable to a chemical reaction becween Zr-4 and A1 023 present at temperatures below 1300*F. This temperature is higher than that which will occur at burnable poison pellet surfaces during Condition II and III occurrences. The reaction between Zr-4 and A1.,0 3 descreed by Idaho Nuclear was obsened to occur rapidly only at temperatures in excess of 2500*F, well above the peak Zr-4 temperatures in the higher-energy fuel rods.
4.0 FUEL SYSTEM DESIGN EVALUATION Discussed in the following paragraphs are the evaluations required for the Batch M reload fuel design. Other evaluations that are not 22
_ . . _ _ _ _ _ _ _ _ _ _ ~_ _ _______ _____________ ________________ _ _
9 affected by the reload fuel design, (such as seismic events under condition II, events under conditions III, and condition IV events) and who'se results are discussed in the USAR or other appropriate documents were not reanalysed and are not discussed. .
4.1 Fuel Cladding Collanse Collapse of fuel rod cladding may occur as a result of the long-term ovality increase due to creep under external pressure on an unsupported length of fuel rod cladding such as occurs in the spring plenum.
Irradiation creep of the Zircaloy-4 cladding is a function of temperature, stress, fare neutron flux (>l Mev) that accentuates the ovality with time. A point of instability may be reached when the ovality reaches a critical value and the cladding flattens completely into the gap.
The design criterion for collapse states that this critical collapse situation must be avoided. The analysis which verifies that the criterion is satisfied through the design lifetime is described below:
(1) High power levels early in life, and the concommitant high clad temperatures and fast flux, combined with achieving the highest design burnup.are the worst combination of events for potential clad collapse. Consequently, the worst rod from the standpoint of clad collapse considerations is a ficticious rod that is assumed to achieve the limiting value of flux redistributions caused by fuel shuffling between cycles.
(2) Fuel rod internal pressure is predicted as a function of time using the FATES 3 computer code (Reference 1 and 2).
Conservative values for cladding dimensions and fuel properties are used to ensure that a minimum internal pressure history is generated.
(3) The resulting FATES 3 internal pressure history is input to the CEPAN collapse model (See Reference 2) along with the cladding fast neutron flux and temperature histories. Conservative values for cladding dimensions are selected (see Table 4.1.A).
The potential for clad collapse is evaluated in the plenum region. Collapse evaluations for clad are based on correction factors developed for observed finite axial gap formation in modern PWR's (See Reference 3).
In accordance with these methods, C-E has performed analytical
. predictions of cladding creep-collapse time for feed batch M for Fort Calhoun Unit 1, Cycle 11. It is cencluded that the collapse resistance of the fuel rods is sufficient to preclude collapse during their design lifetime. This lifetime will not be exceeded during their anticipated operation; specifically rods will not collapse before approximately 42500 EFPH whereas design EF?H is 34000 for 45000 MWD /MTU. These results are presented in Table 4.1.E.
23
... .... l
4.2 Irradiation Induced Dimensional Changes (Crowth)
Fuel assembly length change results from two distinct mechanisms in the Zircaloy guide tubes: irradiation induced grovch and compressive creep. Growth is produced by radiation effects on the l
Zircaloy crystmilfne structure, and causes the guide tubes to elongate. Compressive creep is the permanent reduction in length of the guide tubes in respense to the fuel assembly holddown forces.
Changes in guida tube length affect the fuel assembly engagement with the reactcr internals, and the shoulder gap (the distance
' - between the top of the fuel rods and the bottom of the upper and fitting) . The length change is important in the evaluation of criteria pertaining to each of these aspects of fuel performance.
The overall elongation of a Zircaloy clad fuel rod is due to a combination of the stress-free irradiation growth of the Zircaloy cladding, mechanical interaction between the Co., fuel penets and the Zircaloy cladding and creepdown of the clad 81ng under external coolant pressure. Each of these factors is related to the time of operation through accumulated burnup or fluence. The overan aff act of these influences is an increase in rod length at the higher burnups.
Because the poison rods are shorter than the fuel rods, and are not affected by pellet clad interaction, they are predicted to have greater shoulder gaps than the fuel rods at and-of-life.
The key input parameters for the irradiation induced grovch and creep models are given in Table 4.2.A. The major results of these analyses are given in Table 4.2.3.
4.3 Cladding Strain and Fatisus Analvsis Cladding tensile strain occurs when the fuel pellet or burnable poison penet unrestrained diameter is larger than the inner diameter of the cladding. This win occur when the combination of cladding creepdown and penet svening have closed the diametral gap between the pellets and cladding. The subsequent increase in pellac diameter that produces tensile strain of the cladding can be due to either further irradiation swelling of the pellet material or additional thermal expansion from local power increases. Permaner-(unrecoverable) strain of the cladding takes place if the stress produced in the cladding by the pellet diameter increase exceeds the yield strength of the cladding, or if the stress remains in tLa cladding long enough for creep to occur.
Fatigue is the term applied to the damage which occurs in a material each time it is stressed and unstressed. Repeated application of cyclic stress levels above a certain value, known as the endurance limit, vill eventuany produce a f atigue f ailure. Materials testing is used to establish both the endurance limit and tha critical 2!.
I
4 number of cyclus at given cyclic stress levels above the endurance limit. Methods exist to account for the cumulative damage which occurs when s veral different stress levels are applied to a component during its lifetime.
The analyses to determine the cladding strain and the cumulative fatigue damage were conducted. These analyses show that the unrecoverable circumferantial strain is less than 1% and the cumulative fatigue damage does not exceed 0.8 for all rods.
Specifically, the calculated total strain (elastic and plastic unrecoverable) is later and the predicted cumulativa fatigue damage factor is later at end-of-life.
. The key input parameters for the cladding strain and cumulative fatigue damage analysis are given in Table 4.3.A. The major results of these analyses are given in Table 4.3.3.
4.4 Maximum Fual Rod Internal Pressure The fuel rod internal gas pressure vill increase with burnup as fission gases are released from the fuel matrix and as internal void space is filled by fuel swelling and clad creepdown. In addition, the hot gas pressure is dependent on the operating power level due to relative thermal expansions of the fuel rod components. The-total internal gas pressure, equal to the sum of the partial pressures of the helium fill gas and the released fission gases, can approach values comparable to the reactor coolant system pressure.
4.4.1 Design Methods l
Both fission gas release and fuel rod internal gas pressure are computed using the FATES 3A fuel performance code (Reference 1, 2, and 7). {
4.4.2 Results As a result of the peak pressure analyses the following design criteria have been shown to be satisfied:
(1) The primary stress in the cladding vill not exceed the design stress limit.
25
I (2) The internal gas pressure will not cause the cladding to creep outward from the fuel pellet surface while operating at peak ~
linear heat rates up to the design limit during normal operation. Although the design criterion on m rim m pressure does not require the fuel rod internal pressure to be less than the reactor cooltat system pressure, and the occurence of a positive differentizi pressure would not adversely affect normal operation if appropriate criteria for cladding stress.
strain, and strain race were satisfied, the computed fuel rod internal hot gas pressure remains below coolant pressure values throughoce its lifetime (ie. gas pressure is less than 2100
- psia).
The key input parameters' to the maximum fuel rod internal pressure analysis are given in Table 4.4.A. The key result of this analysis is displayed in Table 4.4.B.
The calculated fuel rod maximum internal hot gas pressure variation with burnup is shcvn in Figure 9.
Figures 10, 11, and 12 depict representative temperatures calculated at three different axial locations along thal fuel rod in the FATIS'IA "
, analysis for = rim m rod internal pressure. ,
~
,. Figure 13 illustrates the calculated gap conductance at each of three axial locations as a function of rod average burnup. Figure 14 shows predicted fission gas release percent from the FATESA run as a function of red average burnup.
Tha gas release is used to coreute rod internal pressures shown in Figure 9.
~
4.5 clad Bovint m
IM
=
,l Extended burnup fuel does not have power peaks near the limiting peak of lower burnup fuel because of its lower reactivity and lover fissile content. Thus, in general, lower burnup assemblies will be at higher power levels and vill be limiting for thermal margin calculaticus. It is therefore reaffirmed that the rod bov vill not li=1t the fuel perfor=ance for assembly average burnups up to 45 C*a'D/ErU.
26 i
[
4.6 Corrosion and Hydrogen Pickun Cladding oxide thicknesses, determined r.etallographically, on -
(-cycle fuel rods (rod average burnups to 55.7 GWD/MTU irradiated in Fort Calhoun reactor became available in 1986 (Reference 5). The new data together with the earlier data presented in CENPD-269-P indicate no significant increase in the rate of corrosion with burnup., Nnreover, the new data confirm the extended-burnup performance of C-E Fuel in Fort Calhoun reactor.
The hydrogen concentration of the fuel cladding after 5-cycle exposure (local burnup N52 GWD/MTU) in Fort Calhoun reactor was estimated to be 150-200 ppa (Reference 6) . At this hydrogen level, the measured high temperature burst ductility of cladding was not performance limiting (Reference 5). It is therefore concluded that hydrogen pickup at 45 GWD/NIU (batch average) burnup will not limit the fuel performance.
4.7 Pellet-Clad Interaction
~
~
In addition, as burnup increases, the capability of the' fuel to reach the power levels needed for PCI f ailure is diminished.
Thisfact(I
,((suggeststhattheoverall probability of PCI f ailures does not increase with rod-averaged burnups up to 52 GWD/MTU range.
4.8 Power to Fuel Centerline Melt The most limiting fuel rod in the core has a power to fuel centerline melt in excess of 22 kw/ft at all times in life.
5.0 Fuel Bundle Assembiv Comcatability 5.1 Hvdraulie Comentabiliev This section provides the basis for hydraulic compatibility for the Batch M C-E fuel with resident Exxon fuel through an evaluation of the impact of design changes made to the Omaha Batch M reload fuel relative to the last batch (Batch G) supplied by C-E to OPPD.
Specifically, the design changes from the Omaha Batch G fuel design to the Batch M design that aff ect the hydraulic perfor=ance are:
- 1. Moving the LE5 plate cicser to the cere support place (CSP) by
(( ]}in,relativetothatforBatchG.
- 2. Chamfering [ ) flow holes in the Batch M LE? place as ce= pared to 84 chamf ered flow holes in the Batch G design.
27
i 3.
Adding ]onthebottom surface o{ f LIF plata.
~
- 4. An increase in the span between the Inconal and first Zircaloy grids from 13.75 in, for Batch G to 14.25 in. for Batch M due to the shortening of the LIF (item 1).
- 5. Reduction of fuel rod length from 136.698 in. for Batch G to 136.675 in. for Batch M.
Analysis shows that the above design changes will result in a negligibly small reduction in the fuel bundla hydraulic resistance
- Lovering the LET place, and incomp[arisonwiththatforBatchG.3anses adding , a small incrasse in resistance which more than offset by the decrease in resistance resulting from more chaafared flow holes in the LEF plata. The 0.023 inch. reduction in rod length has negligible hydraulic impact.
Since the entrance loss coefficiant for the Batch M fuel is essentially the same as Batch G, and the r===4ning fuel components are geometrically the same for both batches, the hydraulic uplift forces and pressure irop characteristics of the Batch M fuel vill be essentially the same as for Batch G. Since the Ezzon fuel bundlas have been shown to be hydr =14-=117 compatible with C-E's Ratch G fuel, and Batch M has essentially the same hydraulic characteristics, the Batch M fuel is hydr =14-='17 compatible with the resident Exxon fuel.
In s::.anary, the hydraulic characteristics for the Batch M fusi are .
comparabia to those for the Batch G fual; thus the Batch M fuel is compatible with fuel supplied by Ezzon.
5.2 Mecha.ical Comoatab111tv This section provides the basis for the statement that the Batch M C-E fusi is comparabia with the resident Exxon fuel. As shown in Figure 4, the relative locations of the upper and lower and fittings and the spacer grid assemblies are at the same elevation. In addition, the cross sectional area at the various elevations is the same . Because the material selection for the components is based on compatability vich existing fuel it can be concluded that the Batch M fuel is mechanically compatible with the fuel supplied by Exxon.
J 28
References
- 1. "C-E Fuel Evaluation Model," CENPD-139-P-A, July 1974. .
- 2. " FATES 3 - Improvements to Fuel Evaluation Model," CEN-161(B)-P, July 1981.
- 3. "CEPAN Method of Analyzing Creep Collapse of' Oval Cladding, Evaluation of Interpellet-Gap Formation and Clad Collapse in Modern PWR Fuel Rods," EPRI NP-3966 Velume 5, April 1985.
4 " Extended Burnup Operation of PWR Fuel," CENPD-269-P, Rev. IP, July 1984.
- 5. A. M. Garde, " Hot-Cell Examination of Extended Burnup fuel from Fort Calhoun," CEND-427, DOE /ET/34030-11, September 1986.
- 6. G. P. Smith, "The Evaluation and Demonstration of Methods for Improved Fuel Utilization," CEND-414, DOE /ET/34010-10, October 1983.
- 7. Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E), " Safety Evaluation of CEN-161 (FATES 3)", March 1983.
l l
l 1
I 29 l
4 Table 1 Omaha Batch M Poisco Rod Assemblies Parameter Batch M Value Poison Rod Drawing No. E-8184-701-110, Rev. 02 Burnable Poison Pellet C-STD-701-014, Rev. 01 Drawing Poison Pellet 0.D. (in) 0.362 3-10 Loading (gnB-10/in) 0.024 Poison lod Length (in) 136.541 Clad 0.D. (in) 0.440 0.384 Clad I.D. (in)
Spacer Tube 0.D. (in) ( )
Spacar Tube I.D. (in) [ ]
Total Spacer Langth (inches) 14.0 Poison Langth (inches) 114 Spacer Tuba Material Zr-4 p
30
Table 2 Clad Physical Properties Rm. Temp (70*F) 670*F 6 6 Modulus of Elasticity (psi) 15.2 x 10 10.2 x 10 i- Poisson's Ratio 0.295 0.250 Thermal Coefficient of Expansion inch / inch *F
-6 -6 Diametral 2.97 x 10 -6 3.19 x 10 -6 Axial 2.97 x 10 3.15 x 10 Yield Strength (tension)(psi)
Ultimate Tensile Strength (psi)
Uniform Elongation % Strain (Unirradiated minimum) _
9 O
3I
l I
Table 3 Data Used for Eatch M Yuel Design TUEL PELLETS .
Nominal Initial Density
. Tolerance on Initial Density Maxisma Pellet Densification Typical Crain Size
~
Pellet OD Tolerance on Outer Diameter Pellet Length Pellet Chamfer ID Pellet Chamfer Height Pellet Dish Depth Pellet Land Diameter Pellet Chamfer Volume Pellet Dished Volume Typical Open Porosity Typical RMS Surface Roughness Sorbed or Indigenous Gas Content U235 Enrichment Tolerance on U235 Enrichment TUEL CottTi Active Fuel Length Tuel Column Dished Volume' Stack Height Density Waf,ght of Uranium FUEL ROD Cladding Material , I Cladding Outer Diameter !
Cladding Internal Diameter l Cladding Thickness, Nominal ,
Cladding Thickness, Minimum Cladding Thickness, Maximum Cladding RMS Internal Surface Roughness Final Annealing Temperature of the Cladding Nominal Cladding Ovality Cladding Ovality Tolerance
?
32
Table 3
~
Data Used for Batch M Fuel Desien Diametral Gap Nominal Diametral Cap, Mini ==
Diametral Gap, Maximum Cladding Length Upper Planua Length Spacer Volume
- Spring Volume End Caps Volume Total Free Volume Fill Gas Pressure, B0i, at STP Fill Gas Composition Cladding Weight End Caps Weight Spacere Weight Spriss Weight Tots 1 Weight Excluding Fuel Column CT.A GUIDE Tt*5ES AND SLEEVES Guide Tube Material Outer Guide Tube OD, Flange Region Outer Guide Tube ID, Flange Region Outer Guide Tube CD Outer Guide Tube ID Total Length Outer Guide Tube Weight ,
Center Guide Tube 0.D., .
Center Guide Tube I.D.,
l Total Length Center Guide Tube Weight l
9 33
9 Table 3 Data Used for Batch M Fuel Design SPACER AND LOWER INCONEL SPACER Grid Materials Outer Strip Thickness
, Outer Strip Height Inner Strip-Thickness Inner Strip Height Maximum Square Size Volume of Spacer Grid Volume of Inconal Grid (Excludir.g Skirt *,
Volume of Grids in Active Core Weight of Spacer Grid by Materini Weight of Inconel Grid by Material Hydraulic Pressure Loss Coefficients, Ncminal
.(Nominal and 1 Cacertainty. -(Referenced to the Assembly In-Reactor Bare Rod 'ilow Area). (Required only if changed from Batch E design.)
POISON ?fLLETS Poison Pellet Loading Poison Pellet Density
- Poison Pellet OD i Poison Pellet Length Poison Pellet Dish Volume Poison Pellet Chamfer volume POISON COLUMN _ )
Active Poison Length Poison Loading in Active Length Stack Height Density Poison Column Weight Poison Column Poison Weight Diametral Cap, Nominal POISON RCD l
Cladding Material l, Cladding OD Cladding ID i Clacding Length -
Cladding RMS Surface Roughness l j
! Cladding Weight l l
t 3 t.
Table 3 Purchaser Data Requirements for Reload Fuel Upper Plenum Length 7.353 Inches Lower Plenum Length 0.0 Inch Total Free Volume 3.546 Cubic Inches
~
Total Weight 1286 Grams FUEL ASSDGLY Total Length (Excludes 2 inch extension 146.333 Inches of Alignment Pins)
Total Weight of Cladding (No Shim Assembly) 93 Kg Total Weight of Zire (No Shim Assembly) 111 Kg Total Weight of Uranium (No Shim Assembly) 378 Kg Total Weight (No Shim Assembly) 535 Kg Total Displaced Volume, Nominal 1.91 Cubic Feet Fuel Rod Growth Clearance, Cold 1.600 Inches Equivalent Dismater of the Coolant 0.04508 Ft.
Passage (s) 2 Assembly In-Reactor Bare Rod Flow Area 0.24492 Ft At Cold Conditions 4 ,g4g Pressure Drop as a Function of Reynolds 3.951 + 26.46 Re 0 53
+ 706 Re b.68
+ 409.4 Re Heat Flux Augmentation Factors See Table 5 Fuel Rod Pitch Tolerance to.005 Inch Fuel Rod Diameter Tolerance to.002 Inch i
a 35
f 4
+
" Table 4 1.ist of Omaha Batch C and H Components, Fuel Drawings and Associated Changes i Summary of Batch C Drawing Numbers Batch H Drawing Numbers Hajor Design Changes Components IIpper End Fitting Center E-20466-701-204 Rev. 04 E-20466-701-204, Rev. 04 None 1.
! and Outer Posto Ilpper End ritting Center E-24066-701-204, Rev. 04 E-24066-701-204, Rev. 04 None 2.
and Outer lloid Down Springs
' 3. lipper End Fitting llold Down E-24066-701-214. Rev. 05 E-24066-701-214 Rev. 05 None l Plate E-STD-701-Il4, Rev. Il None. Both flow plate
- 4. Ilpper End Fitting Flow E-24066-701-214. Rev. 05 ,
- designs have the same Plate l nominal dimensiona.
Minor dif ferences occur u
- in some of the tolerances and other characteristics such as the coplanarity i
requirement for the 5 l bosses. The standard
! drawing is used in all l
CE 14 x 14 fuel bundle
' assemblies l E-24066-701-204, Rev. 04 E-24066-701-204, Rev. 05 None. Only difference is
- 5. Upper End Fitting Assembly the use of a standard 14 x 14 flow plate.
E-STD-701-ll6, Rev. 04 E-STD-701-Il6, Rev. 06 Revision shows the
! 6. Inconel Spacer Crid changes made to the Assembly geometrica of the
! perimeter strips windows and apring tabs.
Upgraded drawing to current CE QADH standards l'
1 l
i
n Table 4(Cont'd.)
List of Omaha Batch C and H Components, Fuel Drawings and Associated Changes i
- Suennary of Components Batch C Drawing Numbers Batch H Drawing Numbers Major Design Changes E-STD-701-il9, Rev. 03 E-STD-701-Il9, Rev. 05 Revised the geometries.
- 7. Inconel Perimeter Strip of the perimeter strips I
windows and spring tabs to enhance the quality l
and performance of the l inconel perimeter strips by making them stronger and less prone to veld li burnout during fabrication.
Upgraded drawing to current CE QADN standards.
Inconel Spacer Strip E-STD-701-018 Rev. 07 E-STD-701-018, Rev. 08 None j u, 8.
" with Arch l None l
4 E-STD-701-017 Rev. 05 1 9. Inconel Spacer Strip E-STD-701-017. Rev. 04 Zire Spacer Crid Assembly E-STD-70!-006, Rev. 06 E-STD-701-006, Rev. 08 Revision shows the 10 changes made to the i geometries of the i
i perimeter strips windows and spring
)
tabs. Ilpgraded drawing to current CE QADH standards.
~
! Revised the geometries II. Zire Perimeter Strip E-STD-701-009, Rev. 06 E-STD-701-009. Rev. 08 of the perimeter strips windows and spring taba to enhance the quality and performance of the 1 Zircaloy perimeter l1, strips by making them 1 stronger and less
]I i
i i
Table 4(Cont'd.) .
.I.ist of Omaha Ratch C and H Components, Fuel Drawings and Associated Changes .
Summary of Components Batch C Drawing Numbers Ratch M Drawing Numbers Major Design Changes l
t I
prone to weld burnout during fabrication.
f UpRraded drawing to current Cg (NWWI standards.
l 1
j Zirc Spacer Strip with Arch E-STD-701-008 Rev. 07 E-STD-701-008, Rev. 08 Revised tolerances and j 12. design information.
] Nominal design dimensions are the same. Ilpgraded drawing to current CE QADM standards u$
E-STD-701-007. Rev. 04 E-STD-701-007. Rev. 05 Revised tolerances
- 13. Zire Spacer Strip and design information.
Nominal design dimensions are the same. Upgraded i drawing to current CE QADM standards.
-1
' I.ower End Fitting E-24066-701-305 Rev. 03 E-8184-701-105. Rev. 01 ,Placed 14.
~ ~
I Reduced height of end fitting from 3.167 1 inches to inches.
Machinea 3/8 inch by 1/4 inch weld slots i
on bottom of post (two per post) in place of weld windows I inch by 13/32 inch located i
3/8 inch shove botton I
of post (two per post).
f .
1
Table 4 (Cont'd.)
1.ist of Omaha Batch C and M Components, Fuel Drawings and Associated Changes Summary of Batch N Drawing Numbers Major Desian Chantes Components Batch C Drawing Numbers D-STD-701-002, Rev. 66 The end cap geometry Feiel Cops Rod End D-STD-701-002, Rev. 04 was changed slightly to 15.
agree with current designs. The end cap envelope remains the same.
18pgraded drawing to current CE QAlti standards.
E-Sl04-701-102, Rev. 02 Reduced overall length Fuel Rod Assembly E-24066-701-302, Rev. Of of fuel rods from t
- 16. l 136.698" to 136.675" I (0.023 inch reduction). !
Revised plenum spring U$ design to increase fuel rod void volsene.
Revised drawing to make C-24066-701-309. Rev. Of C-STD-701-109. Rev. Of it a atendard drawing
- 17. Fuel Pctiet whereby U-235 enrich-ment is specified by H.O.
Weight percent U-235 increased from 3.03I to 3.801.
E-STD-70l-llo, Rev. 05 New drawing. Folson Folson Rod Assembly None rods were not used in 18.
Ratch C.
C-12581-701-514. Rev. 02 New drawing, Folson None rods were not used
- 19. Burnable raison Fellet in Batch C.
1 f .
I .
t Table 4 (Cont'd.)
f List of Omaha Batch C and H Components Fuel Drawinas and Associated Channes l Summary of i,
Batch H Drawing Numbers Hajor Desian Chantes I
componenta Batch C Drawing Numbere E-Sl84-701-le), Rev. 02 Increased length from 20 Center and outer E-24C66-701-203. Rev. 03 top of lower end. fitting l
Culde Tube Assembly to nyper end fittin j
i flow plate by [ inch.
Changed lower end fitting -
l I to guide tube connection.
Belt Eliminated lower nut.
- 21. Lower End Fitting Lower Nut Replaced with[ ]
E-24066-701-203. Bew. 03 D-STD-701-il3, Bew. 03 ~~
Connection Components Locking Disc
-~
Locking Disc E-STD-701-103, Rev. 06 D-STD-70l-ll3, Rev. 02 E
j i
1 1
E-8184-701-.Olt Rev. Of New drawing.Shows
- 22. Fael Bundle Grid None installation require- l Cage Assembly ments prior to rod 1 loading and upper i end fitting installation.
I E-8184-701-101 Re'v. Al See component design
- 23. Fuel Bundle Ammebely E-24066-701-40). Rev. Of changes outlined above.
The dimensional envelope j is changed from 8.190
' inches square to 8.250 inches square.
~
D-8184-701-120 Rev. 01 Not Applicable.
- 24. I. fat of Haterials D-24066-701-420. Rev. 04 l
1
(
t
Table 4.1.A .
Paramecer Name Assumed Value 0.416 (Max, Mid Wall)
Clad Diameter Inch
- Clad Thickness, Inch 0.026 (Min) ovality, Inch 0.00083 Outlet Temperature. *T 646.I*T 1250 Thru 1400 Varies with Exposure Pressure, Psi 21 (In Spring Plenum) fluence (nyt) E> 1.0 Mov 4.07 x 10 Table 4.1.B Minimus Collapse Anticipated End of Batch Life Exnosure Time 42500 ETPH 34000 ETPH for bundle
. M average burnup of 45000 WD/MrU i
j 41
Table 4.2.A Parameter Name Assumed Value Temperature Inlet, *T 539.0
. Temperature Outlet. *F 589.0 - 612.7 (Depends on Case) 21 Estimated EOL Fuel Rod Fluence, n/cm 9.10 x 10 E> .821 Mov Estimated EOL Guide Tube fluence, n/cm 2 8.20 x 10"l E> .821 Mev Guide Tube Material Annealed Zircaloy-4 Fuel Rod Haterial Cold idorked Stress Relief Annealed Zircaloy-4 i
1 j
Table 4.2.3 Batch Parameter Allowable Limit Calculated 21 M Minimus Shoulder Gap 0.0 Inch Limit not reached at 13 x 10 21 Min. Axial clearance 0.0 Inch Limit not reached at 13 x 10 of Upper End Fitting Posts with Fuel Alignment Place Fitting Posts with Fuel A111gnment
- Place 4
l .
1 i
3 42
Table 4.3.A Parameter Name Assumed Value clad Inside Diameter Min., Inch 0.3825 ,
Clad Thickness Max., Inch 0.02975
~
Max. Clad 0.D. Temperature, *T 652 Bundle Design Burnup, m'D/MTU 45000 Standby Pressures, Psia 750 (Min. Pressure Case) 2150 (Max. Pressure Case)
Table 4.3.5 Batch Parameter Allowable Limit Calculated M Strain 1.0% in/in Limit reached (Net Unrecoverable) at 63800 MWD /MTU Peak Rod Burnup Fatigue Damage Factor 0.8 Limit reached at 66400 5'D/MTU Peak Rod Burnup 43
l Table 4.4.A FATES Input for Maximum Design -
, Pressure Analysis Parameter Name Value -
Clad O. D., Inch Clad Thickness, Inch Pellet O. D., Inch
. Active Fuel Length, Inches Fill Gas Pressure, Psia Nominal Initial Fuel Density, %TD Fuel Densification :TD Initial Open void Volume. In 3 CoreAverageFastTjux (E) 1 Mov), n/cm see 2 Coolant Mass velocity, ibs/hr-ft System Coolant Pressure, Psia Core Inlet Temperature. *T Peak Steady State LHGR, Rw/ft Peak Allowable Short Term LHGR, KV/ft E0L Peak Pellet Burnup, MWD /MTU ICL Rod Average Burnup, MWD /MIT Table 4.4.3 Fates Output for Maximum Design l
Pressure Analysis Parameter Name Value r--
Maximum Predicted Internal Gas Pressure, Psia
,l l +
FIGURE 1 4
COMPARISON OF OMAHA LOWER END FITTINGS i * ,.
i 1
I i
P 1,
- l 4
I e 4
i i
1
(
=W k#
f i '
1 9
2 I 6 l
i 1 ,
1 4
4 t
l l
i i
1
.i 4 l
i
-o. - - . - . ,
e e
, e
- t 1
E i 4 l
A f .
- s i
C O
t ; -
nm O
i WQ .
i C3 w
C I ww
, a E i m i
! o -
l
~
k l
i I
I I
I 46
FIGURE 3 OMAHA BATCH M FUEL BUNDLE DESIGN
'$d I
s I I
. . i i
i
. i il i I
. 1
=
i ; -
I i I
.& , f ll 1 a a'a n n'= s#= --
<- m,'svv i n M a r 0203020s030sDE -
SmnmmmenanAnv
)O.AA)[b. lib 4 23EDCOCO20C030 INIMDDY y.L- l ymw 0503030303030Z' vnn, ow nrH l_ __
C. 2 0 2 0 3 .0 2 0 2 0 3 0 2 I e e e e se o e e A
l l w$
[; U-~
i i i
! V V
(
47 l
l l
. = _ . -
l l
- FIGURE 4 COMPARISON OF SPACER GRID LOCATIONS OF C E's AND EXXON RELOAD FUEL BUNOLE ASSEMBLIES bb i i I b bbii 6 u I bbb i il y l ' n
, l l l l I I i l l I I t i ! l l 1 i e l I IP I I I
l l 141.821 y[
O~ ,,,l. 141.021 #, , y 137 I ' 137 i 137 i a a
I J 1 f P
' ians i I , tani i , inn 2 __
- esp *"
r' P
- 1. I e- in.3 I i'~ ' I in.3 i f e 1
- if 4 1P
- s, 1P
, - r- ,,
I 87.158 87.150 87.157 l f 1
- l l 6 I ?
1 fii =_
F ,
a
r' .
i
a
ir il it 1P 1P
.qP 53.525 53.525
, 53.332 8
I I I I i__ .
1P y 1r If IP if ' '
' ' ' 38.713 38.713 38.729 ,
1.375 1.375 v n i
I i i I__ a i i n
i 1P i 1P 17 o a e ,
' 19J 13.807 19.9 Ii .
o INCONEL o
, ,, INCONEL
" GRID o GRID l.543
(
' 8.964 8.427 I ' o i i n .
l y 9 i o . . l l
l
~
l 2.737 ,j ".1d i i 3.187 _
l l
l 2.746
- i i C E BATCH G FUEL C E BATCH M FUEL l EXXON FUEL I
48 j
FIGURE 5 FUEL RCD UPPER #
END CAP SPRING I
SPACER E
'eN A
~
FUEL PELLETS ACTIVE
- FUEL LENGTH
.m 128 FUEL CLADDING _
t SPACER Y
\ _
- rt:
LOWER "
END CAP 49
O e
FIGURE 6 BURNA8LE POISON ROD 9
e h
i i !
i e
i I I
i 50
0 F1. CALHOUN RADIAL FALLOFF CURVE i -
RGURE 7 1
i
! l i
i i
I i
! i 51
... . . _ . . I
O e
b e
O AXIAL POWER DISTRIBUTION ENVELOPES RGURE 8
=
l e
4
,T i
I i
! \ '
i .1 i
l i
i
. I i
I i I l
. .s A
52
\
4 e'
e e
e e
FT. CALHOUN MAXIMUM PRESSURE
- FIGURE 9 ,
Y t
, i f
I l
l I
I l
1 l
l l
l I
3 e
l l
4 I
i 53 l _ - _ _ .
l e
s' l
)
TEMPERATURE AT PEAK POWER NODE
, F1GURE 10 i
e-1 i
l K
I I
l l
l l
5 34
.a-M wa & 4 -
w- ,-M- A 4u S'
e a
L e
i TEMPERATURE AT AVERAGE POWER NODE
, , FIGURE 11 i
i E
e i
6 53 e - - , - - , - - - , ., , - , , . , . - , - - , - . -
a m .a A: ,_ -.: ..as a _A s 5_ m...- _m4 + . _ . - m m - 1.. __ ..a.m_#;m.. .2._ _ --+ .,% -. 2 , . mu .
e
- P l
k '
I I
TEMPERATURE AT LOW END NODE
- MCURE 12 t
j l
'l i
f I
t 1
i 56 ,
p GAP CONDUCTANCE
. RGURE 13 e
4 57
4 I
FISSION GAS RELEASE PERCENT MGURE 14 i
I 58
._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _