ML17252B198
| ML17252B198 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/20/1973 |
| From: | Commonwealth Edison Co |
| To: | US Atomic Energy Commission (AEC) |
| References | |
| Download: ML17252B198 (71) | |
Text
£,
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SYPPLEMENT D TO DRESDEN 3 SECOND RELOAD LICENSE SUBMITTAL
. ~ : '
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, I
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CONTENTS 6.1 6.2*
6.3*
6.4*
6.5*
6.6*
6.7*
. 6.8*
6.9*.
6.10*
6.11 6.12 6.13
'QUESTIONS AND RESPONSES REFERRING TO ATTACHMENT A Dresden* 3 Nuclear Power Station Second Reload License Submittal, September 1973
- Not contained - to be submitted the week of December 31, 1973 *
\\.o o.11 QUESTIONS Provi,de a schedule for the submittal of promised information (Section 6.2.3.2) which evaluates the thermal-hydraulic s~ability of the 8x8 assemblies and the.core with a mixture of 8x8 and 7x7 assemblies.
RESPONSE
This information has been provided in "Dresden-3 Supplement B to
- Second Reload License Submittal, AEC Docket 50-249, 11 dated December 6, 1973.
6.1 QUEST! ON In_ reter~nce ~~_Figures _6-4 and 6-5 discus_s the variation of the scram - -
reactivity function with burnup throughout the cycle.
Do these figures*
represent the worst case during the cycle?
Is the proposed technical specification scram time of your letter of August 1, 1973 used?
Exp.1ain
- the differences figures curves 6-4 and 6-5 and Figures 6 and 7 of your letter of September 14, 1973 concerning maximum _allowable in-sequence control rod worth.
RESPONSE.
The scram reactivity feedback function experienced during a control rod drop accident will vary slightly during the cycle from that shown in Figures 6-4 and 6-5.
The curves labeled "DRESDEN 3 PERFO~NCE" repre-sent middle-of-cycle conditions with actual expected scram times.
Scram reactivity feedback insertion rate is not strongly dependent upon core exposure but by the end of the cycle the relative scram worth as a
--fun-ction-*of insertion distance may decrease slightly as shown in Table 6.1 below.
The total scram worth will, however, increase because the number of rods that must be withdra1.n at the start of the RDA to achieve criticality increases. This 0 f7ect is also illustrated in Table 6.1.
The "DRESDEN-3 PERFGRr*iANCE" curves shown in Figures 6-4 *and 6-5 are those asso' 1ated with the drop of the maximum worth control rod at the mo~erator conditions noted.
\\.
Table* 6.1 Scram Response at Hot Startup Conditions Scram Bank Figure 6-5 {~MOC3)
EOC-3 Position, feet Inserted 6k 6k/6k total 6k 6k/6k total 0 - 6 0
0 0
0 9
.007 0.10
.009
.07 10.5
.031 0.44
.042
.33 12
.071 1.0
.125 1.0 The proposed technical specification scram time of the August 1, 1973 letter is used for the 280 cal/gm bounding curves.
In regard to the differences in.the curve showing the scram reactivity function the following applies:
- 1.
In both cases, one when the "typical plant" performance is shown and the other when Dresden 3 performance is shown, two sets of curves are presented.
One is for the cold startup condition and the other i~ for the hot startup condition.
In both cases the actual p, :.*nt performance compared to the bound-ing condition is shown.
- 2.
There is a difference between "typical plant" performance and Dresden 3 performance.
Each plant, in fact each core loading would show a difference from "Typical Plant" performance.
The
- significant measurement is that actual plant performance does not ~xceed the bounding curve of the scra1J1 reactiyity function..
Maximum in-sequence rod worth has been determined by calculation considering the limit of 280 cal/gm peak fuel energy content.
This is calculated to occur at the bounding curve.
- 3.
The average measured scram time at this plant is used for the "Dresden 3 Performance" curves.
6.12 QUESTION Provide a schedule for the submittal of the evaluation of abnormal transients, such as core coolant flow decrease, "!"Jr which "confirmatory calculations are currently in progress 11 (Section 6.2.3.2).
RESPONSE
This information has been provided in "Dresden-3 Supplement B to Second Re 1 oad License Submittal, AEC Docket 50-249," dated December 6, 1973.
- 6. 13. lli!l_)TiON Provide specific comparisons of the 8x8 and 7x7 assemblies which demon-strate that the "conservative fuel type" as stated in Section 6.1.3.1 of the 03 submittal has been used in the analysis of each abnormal.
operational transient and the nuclear system pressur~ increase transients in particular.
RESPONSE
The licensing topical (NEDO 10802) describing the transient analysis methods used for the simulation of these events shows the key importance of the "average" fuel element in all total-core transients like the pre~sure increase events. Since the projected fuel load is ~till dominated by 7x7 fuel, this fuel tYPe was used to prepare the tran-sients analyzed for the reload.
Physics characteristics used for the averaged core do reflect the mixture of fuel types and individual channel thermal hydraulic conditions. Comparison pressurization transients have been made in which all 8x8 fuel geometry was assumed.
In all cases, results were similar with the all 8x8 peak pressures
.. slightly less (5-10 psi) than the 7x7 dominated simulation assumed for the license submittal.
_?imil ar considerations were co.vered for other key transients, such as the recirculation flow coastdown cases.
In each of those situa-tions the hottest channel was simulated to be either 7x7 or 8x8 starting from limitinq {1.9) MCHFR con*!itions using the detailed, multi-node thermal hydraulic tools reo*J1arly applied for this evaluation.
The 8x8 bundle consistantly she*.:.::.: slightly lower dynamic thermal margin and is therefore the source of the minimum MCHFR values given. The 7x7 bundle was better by less than 0.1 for the pump trip and seizure cases st~d~~~.
The transient safety analyses provided are therefore presented as con-servatively bounding the mixed fuel combinations considered for this reload~
CONTENTS 3.1*
3.2 3.3 3.4 3.5 4.1 4.2 4.3 4.4 4.5 5.1 5.2 5.3 5.4 5.5
- )
- 6 5.7 QUE-STIONS AND RESPONSES REFERRING TO ATTACHMENT B NED0-20103 General Design Infonnation for General Electric Boiling Water Rea-tor Reload Fuel Conmencing in Spring 1974
- Not contained - to be submitted.
- 3. 2 QUESTION
- For tables 3-3, 3-4, 3-5 and 3-6 of NED0-~0103, provide the dimensio~s of the edge chamfer and end dish of the pellets in each type of rod.
For tables*3-3 and 3-5 provide the mean density of the
.. pellets in each type of rod.
For table 3-4, provide the fuel rod diameter, clad thickness, pellet-to-clad gap, active fuel length, plenum volume, and maximum linear heat generation rate for each type of rod.
RESPONSE
The fuel rod diameter, clad thickness, pellet-to-clad gap, active fuel length, plenum volume, and maximum linear heat generation rate for each type of rod in Table 3-4 is provided in Table 3-3 with the excep-tion of Dresden 1 Type 1 and JPDR fuel.
For these fuel types, the requested parameters have previously been documented in Table 4.1 of
- NED0-10173.
11Current State of KnO'rledge-High Performance BWR Zircaloy-Clad uo2 Fuel 11
, May 1970. There is n-: fuel type with chamfers in the referenced tables of NED0-20103.
The nominal theoretical density of all fuel rods or capsul~s in Tables 3-5 and 3-6 is specified in the titles of those tables. All fuel types in Tables 3-3 and 3-4.employed 95% T.D. uo2 pellet fuel with the exception of Dresden 1 Type l which was 96% T.D.
The dishing of fuel rods in Tables 3-3 and 3-4 was as follows:
2% dish.
for KRB Type KD, Oyster Creek, and Nine Mile Point; 3% dish for Garigliano Types SA and SB, Humboldt Type III, Consumers Types EG and F,
. Tarapur 1 and 2, Tsuruga, Fukushirna-1, Millstone and Nuclenor; 3.5%
dish for Consumers Type E, Dresden-2 initial core and reload, Monticello, Dresden. 3 and KKM.
All other fuel types in Tables 3-3 and 3-4 were not
- dished.
For Table 3-5, the D-50 rods have 5% dish and 3 rods out of 21 in GE - Halden have 3% dish.
For Table 3-6, capsules Band E have 3%
dish.
> f(,;
'I*
3.3 QUESTION Provide the basis for the statement that the flow-induced fuel rod 11vibrational amplitude 11 does not exceed 0.002 inch (Section. 3.2.9).
Describe the tests and analyses of flow-induced vibration in an 8x8 assembly which have been performed.
RESPONSE
The flow induced vibration design analysis is based on test data which encompasses the expected range of conditions for the 8x8 reload fuel.
The empirical correlations derived from these data predict a vibrational amplitude of 0.7 mils.
The test information and data are contained in the following report:
Quinn, E. P.,
11Vibration of Fuel Rods in Parallel Flow, 11 GEAP-4059, July 1962.
I 3.4 QUESTION Describe the tests of BxB fuel assemblies mentioned in Section 3.2.10 which "have been conducted both out of reactor as we 11 as in reactor.
11 Provide the results of "all tests and post-irradiation examinations which have indicated that fretting corrosion does not occur.
RESPONSE
General Electric has maintained a continuing ~uel surveillance program which has included the effect of other fuel assembly components on the fuel rods.
. These inspections have been performed on a large quantity of fuel rods at a number of *operating reactors and have utilized site non-destructive test (NOT) methods including eddy current (E/C) and *detailed visual examinations.
Spacer contact regions on fuel rods, in particular, have been examined on a routine basis to provide informational feedback to the design process.
Within this considerable experience, fretting corrosion has not been observed.
The preponderance of inspection experience has been on two basic spacer designs currently in operating reactors. The first is the 7x7 stainless.
steel spacer which ~ricorporates a wire grid and Inconel iprings as fuel rod contact points.
These spacers were the standard design for the initial cores of BWR-2's and some BWR-3's.
The secc~d type was the 7x7 zircaloy spacer which eliminated the wire grid and ;~:orporated a larger number of Inconel springs which acted as f:~c contact points. This design was employed in the BWR-3's, BWR-4's, and BWR-S's and in later reloads.
The BxB spacer is basicc.~ly identical to the zircaloy 7x7 spacer except for the larger number of rr.ds and the corresponding scaling down of dimensions to accommodate the smaller 8x8 rods.
The fuel rod contact points and contact loads, however, are essentially the same as the 7x7 spacer, and therefore can be expected to perform similarly.
The site inspection methods used in the examinations include continuous eddy current to locate discontinuities in the fu~l cladding, and visual examin0tion to.characterize the nature of defects.
The eddy current is capable of detecting any detrimental effects of the spacers on. fuel rods such as fretting corrosion, but also d~tects surface artifacts such as scratch'es, uneven oxide film or crud buildup.
Therefore, eddy current methods
'*'I j~~.-~..:......
utilized in the field have been used in conjunction with visual exa~inations to ident.;fy any eddy current indications at spacer locations. The visual examinations have be.en performed using a horesc_ope wi_th appropriate magnificatfon to closely examine the exterior surface of fuel rods.
Based on site inspection results, fretting corrosion has not been* observed with the G.E. fuel rod spacer design.
Almost. 15,000 fuel rods from ~ssemblies utilizing the older stainless steel spacer design have been eddy ~urrent
- tested and over 600 of'these!with eddy current indications at spacer locations have been visually inspected.
Most of these indications were observed to be insignificant clad surface conditions such as crud buildup, uneven oxide, scratches in combination with superficial spacer contact marks.
No indication of fretting corrosion was observed even though some of these assemblies have achieved exposure as high as 12,000 MWD/st.
Approximately 12,000 of the.rods from assemblies employing the newer zircaloy spacers have been inspected with eddy current methods and almost 150 rods visually inspected.
No indication of fretting corrosion has been observed on these rods.
Assemblies from which these rods were taken have achieved exposures as high as 7000 MWO/st.
In addition, data has been obtained from two pilot assemblies incorporating the zircaloy spacers which have operated beyond 13,000 MWD/st with no visible indication of fretting corrosion.
Operating experience on produ~+.;~~ bxB fuel spacers is not available since the first batches are i1ut scheduled to be loaded until early 1974.
- However, the 8x8 s~~r~r design is sufficiently close to the zircaloy 7x7 spacer that cheir performances are expected to be similar.
'Ali8t'fn'I). -* -- ~*
- 3. 5 QUESTION Describt:. the post-irradiation surveillance program planned for the 8x8 reload assembly.
Describe the proposed tests and 1nspection-s, the number of rods and assemblies involved and the time-in-reactor of the assemblies.
RESPONSE
An overall fuel surveillance program has been aggressively pursued by G.E.
to monitor the performance of production fuel and to obtain information on I
specifi*c fuel designs. The.program has spec~fically included post-irradiation exa_mination of the lead exposure fuel of specific designs, both at the site and in the hot cell. At site inspections over the past four years, over 27,000 rods have been non-destructively tested employing both ultrasonic.
and eddy current methods, and over 750 rods visually examined with a borescope.
Over 150 production fuel rods, in addition to developmental fue 1 rods, have been returned to the hot cell for various deta i1 ed examinations which have included visual examination under magnification, cladding leak check, fission gas analysis, clad profilometry, neutrography, gamma scan, eddy current, fuel isotopic analysis, and metallography.
As a new confi"guration, the 8x8 fuel design will be incorporated into the fuel surveillance program with the ca~didate bundles for inspection selected from the lead exposure batches. Sinse the first 8x8 reload batches are scheduled for insertion in early 1974, and reactors typically operate on an annual cycle, the earliest possible inspection could not be performed unti1 early 1975. This operation for approximately a year would result in bundle average exposures in the range of 4000 - 5000 MWD/st.
Inspection of the 8x8 will be performed at the earliest opportunity with the degree*
of inspection and number of bundles inspected dependent on the availability of fuel discharged from the core and on the outage plans of the utilities.
Inspection methods can range from radiochemical evaluation of bundle performance to a general visual examination of the bundle periphery to a detailed NOT examination of all fueled rods in a bundle coupled with selected visual inspection.
Fuel rods would be returned t6 the hot cell for more detailed examinations if deemed appropriate by the site examination.
- 4. l.
QUESTION Demonstrate that the referenced CHF correlation is applicable to an 8x8 assembly.
Compare the available data from full scale tests of 8x8 bundles to the predictions based 6n thi~ referenced correlatiori.-
From the test data which most closely approximates the conditions in the hot assembly at normal full power operation, estimate the bundle power which will produce the onset of transition from n.ucleate boiling.
Compare the test parameters with expected operating tonditions (e.g.
bundle power and flow, inlet enthalpy, axial and local power peaking factors).
RESPONSE
The Hench-Levy Critical Heat Flux Correlation was introduced in 1966, and its basis and fonnulation was presented to the AEC in "Design Basis for Critical Heat Flux Condition in Boiling Water Reactors."
The range of significant parameters used in the collection of data for the correlation arc presl.:111.ed in Ti!blc 1.
The test data, coupled 1*Jith a multichnnnel hydraulic model.to *predict the local coolant behavior in mrre complex geometrics than those tested, fonned. the critical heat flux design basis.
Since its initial introduction in 1966, the Hench-Levy Critical Heat Flux corre 1 at ion has been accepted by the AEC as a 1 i cens i ng basis* for the evaluation of thermal margin for a variety of boiling water reactor.
fuel geometries.
A tabulation of principal geometric parameters for these lattice configurations.. appears in Table 2.
Also shown in Table 2 are comparable parameters for the 8x8 assembly.
Use of the Hench-Levy correlation was continued for recent submittals covering 8x8 reload fuel, because the geometri~al, thennal, ~nd hydraulic parameters are similar to those of the fuel in plants previously licensed under this thermal design basis, and well within the range of previous applications.
Reactor operating* conditi~ns fo~ the 8x8 fuel in terms of pressure, mass flux, and* quality do not differ from those previously experienced.
TABLE l HENCH*LEVY CORREL/\\TION TEST PARN1ETERS l.
Multirod Geometry Heated length Hydraulic Diameter Rod*to-Rod Spacing Rod-to-Channel Spacing Pressure Flow Rate Ste.am Quality Heating uistribution
{ *'
Four-and Nine-Rod in Square Array 36, 45, 48, 60 inches 0.324-0.485 inches 0.060-0.187 inches 0.060-0.135 inches 600-1450 psia 0.2-l*.6xlo6 lb/hr-ft2 0-0.6 Uniform and. Increased Uea ting in Corner Rod or Central Rod
~+~*sw~arfii
.. _18.* *r:ii~lilli1'1...
"'11~'Willi...,..............
J
-~ r *=~~ #% 5.... \\iitllllli4'i.V.""*,4*;,,;,,.-;*-,.-...:~i.:.:...--:*.*--....... -.. - *. :-*** ****--***-... **:~*....:-:-1.*-***.,........ _.,..~..,,.-... -.* --..... _. ____....._ __
r
\\J T/\\BLE 2-GE-BWR FUEL ASSEMBLY PARJ\\METERS.
Geometry 6x6 7x7 9x9 llxll 8x8 Heated Length, inches
- 17. 5-109 79.. 144.
70
. 70 144 Hydraulic Diameter,
.453-.567
.494-.576
.499
.497
. 516
- inches Rod-to-Rod Spacing~
- 134-. 177
.145-.175
- 145
.128
.147 inches Rod-to-Channel Spacing,
.087-.140
- 135-. 144
~162
.140
. 153 inches
- ...:.... **-~.....
~-,,....._.._......,,.-~~
............ ~-._.... ---~-r.- ::.*--*****..............,c.:t* "1.:'tA....,_.~... ***-~*.. -.-- _ __..**----.__....!I" ___.,_
j,
- \\
I I
.i 4.2 QUESTION Describe the methods used to calculate the steady state flow distribution in terms of mass velocity between 8x8 and 7x7 assemblies.
Estimate the error in calculating the flow in an 8x8 assembly relative to a 7x7 assembly.
Provide the basis for this estimate.
RESPONSE
The methods used to calculate the steady state flow distribution in a core of 7x7 and 8x8 fuel assemblies are similar to those used throughout the nuclear power industry.
Core the~mal-hydraulic analysis is performed with the aid of a digital co~puter program. This program models the reactor core as several parallel flow paths usins a hydraulic description in each path of the orifices, lower tie-~late, fuel rods, fuel rod spacers, upper tie-plate, fuel \\.;1annel and the core bypass flow.
The basic assumption us~d by the code in performing the hydraulic analysis is that the flow
- entering ~h~ core will divide itself b~tween the fu~l bundles and the bypass fl ow paths such that each assembly and bypass fl ow path experience the same pressure drop.
The code can handle 12 fuel channel types and 10 types of bypass flow paths. Typically each fuel type would be modeled using one 11hot 11 channel and the balance as average channels in each orifice zone.
The group of average channels would ciperate at the average power of each fuel type and define the share of core flow that each fuel type will receive.
The hot channels then examine the most limiting assembly of each fuel type.
In performing the hydraulic
- calculations on each channel type, the orifice, lower tie-~late, fuel rod sp~cers, and u~per tie-plate are hydra~lically represented as being separate *. distinct local losses of zero thickness.
The fuel channel cross-section is represented by a square section with enclosed area equ~l
_J 4.2 Question (Continued) to the unrodded cross-sectional area of the actual fuel channel.
The fuel channel assembly consists of three bfsic axial regions.
The first and most important is the active fuel region.
For a 7x7 fuel assembly, the active fuel region would consist of 49 fuel rods and 7 fuel rod spacers. The 8x8 assembly consists of 63 fuel rods and 1 non-fueled rod with 7 fuel rod spacers. The second region is the non-fueled region consisting*of non~fueled rods and the upper tie-plate. The third region represents the un-rodded portion of the fuel channel above the upper tie-plate. The active fuel region is considered in 24 independent axial segments or nodes over which fuel thermal properties are assumed constant and coolant properties are assumed to vary 1 inearly. *
- The bypass flow paths considered are described in the following table and shown in Figure 4.2-1.
Flow Path Description la. Between Fuel Support and the Control Rod Guide Tube {Upper Path) lb. Between Fuel Support and the.Control Rod Guide Tube {Lower Path)
- 2.
Between Core Plate and the Control Rod Guide Tube Driving Pressure Core Plate Differential Core Plate Differential Core Plate Differential Number of Paths Orie/Contra l Rod_
One/Control Rod One/Control Rod
4.2 Question (Continued)
Flow Path Description
- 3.
Between Core Support.
and the In-Core Support Instrument Guide Tube
- 4. Between Core Plate and Shroud
- 5.
Between Control Rod Guide Tube and Control Rod.Drive Housing
- 6.
Between Fuel Support and Lower Tie-Plate
- 7. Control Rod Drive Cool'ant ro Between Fuel Channel and Lower Tie-Plate
- 9.
Holes in Core Plate Driving Pressure Core Plate'.
. Differential Core Plate Differentia 1 Core Plate Differential Channel Wall Diff-erential Plus Lower Tie-Plate Differ-Number of Paths One/Instrume~t One One/Contra 1.Rod ential
- One/Channel
- lndependent of Core Channel Wall Differential Core Plate Differential One/Control Rod One/Channel Plant Dependent Due to the large flow area; the pressure drop in the*bypass region above the core plate differential pressure and the bypass region elevation head is equal to the core differential pressure.
4.2 Question (Continued)
The computer program iterates o~ flow through each flow path (fuel assemblies and bypass paths) until the total differential pressure (plenum to ~lenum) across each path is equal, and the sum of the flows through each path equals the total core fiow.
- The total core ~low less the control rod cooling flow enters the lower plenum through the jet pumps. A fraction of this passes through the various bypass paths.* *The remainder passes through the orifice in the fuel support (experiencing a pressure 1 oss) where more fl ow is 1 ost through the fit-up between the fuel support and the lower tie-plate into the bypass region.
The majority of the flow continues through the lower tie-plate (experiencing a pressure loss) where some flow is lost through the flow path defined by the fuel channel and lower tie-plate into the bypass region.
In the 8x8 assembly, a small portion of the in-channel flow enters the non-fueled rod through three orifice holes just above the lower tie-plate. This flow, nonnally referred to.as the water-rod flow, remixes with the active coolant channel flow below the upper tie-plate.
The flow through the bypas~ 71uw paths are expressed by the form:
W = C1AP112 + C2~~r~ + C3AP2.
The co~fficients for flow through major paths, such as that through the finger springs, are based on full scale.test data.
The coefficients for the other paths were determined by use of "Handbook of Hydraulic Resistance", AEC-TR-6630, by I. E. Idel 'chek in conjunction with nominal assembly dimensions.
The co~relations used in determining the core pressure drop for single and two-phase flow are discussed in.
answer to other questions.
Within the fuel assembly, heat balances on the active coolant are performed nodally.
Fluid properties are expressed as the b~ndle average at the particular node of interest and arc based on 1967 International
.StandaFd Steam-Water Properties.
In evaluating fluid properties a constant pressure model is u~ed.
--.'*---...--.-w.-..--ir..-...--~-~--~w -..-
- 4.2 Question (Continued)
The core power is di~ided into two parts:
An active coolant power and a bypass fl ow power.
The bypass fl ow is heated by ncutron-sl ow1ng down and gamma heating in the water c:*1d by heat transfer through the channel walls.
Heat is also transferred to the leakage flow from structura) and control elements which are themselves heated by gamma absorption and by n,a reaction* in the control material.
The fraction of total reactor power deposited in the bypass region is very nearly 2%.
A similar phenomenon occurs with the fuel bundle to the active coolant and the water rod flows.
The net effect is that... 96% of the ctire power is conducted.through the fuel cladding and appears as
- heat flux.
The power is allocated to the individual fuel bundles using a relative power factor. The power distribution along the length of the fuel bundle is specified with axial power factors which distribute the bundle's power amongst the 24 axial nodes.
A nodal local power or.
peaking factor is used to establish the peak heat flux at each nodal location.
The relative (radi11) and axial power distribution~ when used with the bundle flow determine the axial coolant property distribution resulting in sufficient information to calculate the pressure drop components within each fuel assembly type.
Once the equal pressure drop criterion has been satisfied, the CHF calculations are performed by applying the correlation expressed in "Design Basis for Critical Heat Flux Condition in Boi 11 rig Water Reactors,
11 at E'.ach axial node for each fuel type.
In applying the ~bove methods to*core design, the number of bundles (for a specified core thermal power) and b~ndle geometry (8x8, rod diameter, etc.) are selected based on power density and linear heat 'generation rate limits. The core design pressure is determined by the
4.2 Question (continued)
_ required turbine throttle pressure, the_ steam line pressure drop, steam dryer pressure drop, and steam separator pressure drop.
The core inlet enthalpy is determined by reactor and turbine heat balances.
The results of applying these method~ and specifications are:
- 1. Flow for each bundle type,
- 2.
Flow for each bypass path,
- 3. Core pressure drop,
- 4.
Fluid property axial distribution for each bundle type,
- 5.
Nodal CHFR calculations for each bundle type.
The effects of uncertaintie~ in design values and on calculation~l results are reflected in the required thermal limits (MCHFR>l.9).
A sensitivity study of design ~arameters has been performed on an 8x8 lattice. The results of this study are presented in Table 4.2-1.
Similar results would be expected for a 7x7 lattice. As shown by this study, applying gross changes in design values (which for the most part are verified by tests) results i:1 insignificant changes in MCHFR.
The methods just described are used to determine the steady state flow distribution b~tween 'all fuel bundles in the core; 7x7 or ~xB. In both assemblies, the basic approach is the same and the cbr~ flow distri-bution is based on a calcu~ated pressure drop across the fuel assembly.
Since similar pressure drop tests have been conducted for both 7x7 and 8x8 assembli~s, there would be little difference between our ability to predict the flow distribution between the two.
As has been demonstrated above, the sensitivity of the fuel bundles thermal margin to the design parameters used is relatively low.
.
- I
.)1.,ltl.:1"11\\11\\, VI 1*,1.1\\l,IUI\\ /\\J.>l.l"1l1l. I
- )1111:*11111.1 1111. l.1./\\h/ll*L. 1;L\\1.1 111111.J I.ow er Tic Plate 7
\\
. ~*- -.Or1vc llou~ing NOTE:
Peri pher.,1
- ruel s11ppot*l::. arc \\*1cl dell into the core $Upport plufo. for Uwse Chunncl bun ell c~, pll th numbers 1, 2, 5 an cJ I do no t exist.
- 8.
6 Core Support In-Core Guide Tube 4
Control Rod Guide Tube Shroud
- 1. Contra l Rod Guide Tube-Fuel Support
- 2. Contro 1 Rod Gui de Tube-Core Support Pl u ti
- 3. Core Support Platc-Incore Guide Tube
- 4. Core Support Pl ll te-Sh roud
- 5. Control Rod Guide. Tube-Drive llousing
- 6. rucl Support-Lower Tic Plate
- 7. Con tro 1 Roc.l Ori ve Coo 1 i ng \\*/I) tcr G. Channcl-l.o*.. 1cr Tic Pl11tc
.9. Sup po r t P 1 u t e llo 1 c s r----*
TABLE 4.2-1 Hot Channel Ave Olannel
- 6. Flow Flow Flow C:;.se. res cri p:i on LB/HR MOiFR LB/HR psid From Base 1
Bc..se Case 115,080. 1.937 127,628 18.68 ------
2 E=:::c gc..1.eo\\;S 2;5.Local Less :Multiplier 114,012----1.918 ' 127,350 19.23. -1068 3
20 % I::crease in Friction 114,099 1.920 127,414 19.44 981 2~% I~~rea.Se in Spacer Loss 114,631 1.930 127,483 19.28
- 449 5
5 ~*:ils Cn:d on Fuel Rods 114,320 1.923 127,418 l9o55
- i60 6
10% Decrease i..~ Central Orifice K
- '.15,614 1.944:
127,492 19.41
+ 534 7
44~ Decrease in Bypass Fla..;
120,668 2.001
- 133,544
- 19. 85. +5588
--- --~~
4.3 QUESTION Explain the difference between the tests "performed in single-phase water to calibrate the orifice an~ tie plate and in both single-and two-phase flow to arriv~ at best-fit design values for spacer and upper tie plate pressure drop (Section 4.1.1.2 of NED0-20103) and the full scale 8x8 tests "performed to determine the local loss coefficients for upper and lower tie plates and fuel rod spacers.". Compare the results of these tests to each other and to the pressure drop at various flows and powers calculated using the standard design method.
RESPONSE
As noted in Section 4.1.. 1.2, single phase and two phase hydraulic parameters are correlated by prototype (64-rod bundle) flow tests. The range of the test data for development of the correlations is summarized in table 4.3-1. The data cover a wide range of test conditions, including the range of operating parameters for 8x8 reload fuel.
Calculations have been performed using the standard hydraulic
. design code to compare.the calculated bundle pressure drop to the measured bundle pressure drop.
Figure 4.3-1 provides such a comparison of bundle pressure drop over the assembly heated length.
I TABLE 4. 3-1.
Measured Parameter Adiabatic Tests:
Spacer single phase loss coeff.
Lower tie plate + orifice single phase Loss coeff.
- Upper tie plate single phase 1oss coeff.
Upper tie plate friction factor Test Conditions T = 100 to 500°F Spacer two phase loss coeff.
Two phase friction multiplier Diabati<: Tests:
Heated bundle pressure drop
- ReynoldsNumber P = 800 to 1400 psia G = 0.5xl06 to l.5xl06 lb/h-ft2 X = O to 40%
P = 800 to 1400 psia G = 0.5xl06 to l.5xl06 lb/hr-ft2
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4.4 QUESTION For the hot 8x8 and 7x7 assemblie~. provide the flow rate, bundle power~ axial and local peaking and exit void fraction at normal full power operation.
RESPONSE
The requested information was provided in Section 4.3 of the Dresden Unit 3 Second Reload Submittal, AEC Docket 50-249, September 14, 1973.
- 4. 5 QUESTION Provide the design correlations, including all constants, used to calculate the friction factor, two-phaie fri~tion multiplier and two-phase local multi pl i.er described in Sections 4. l. L l and 4.1. l.2 of NE00-20103.
RESPONSE
Proprietary infontlation containe~ in separate submittal.
Th~ two-phase friction pressure drop, tiPf, is calculated by w 2 f L
.2
- . d ~ ~ 29('f
- D,.; Ac~ c/)TPF Proprietary infonnation contained in separate submittal.
- 1*;;~;+;41;*
l§~?fl*;,;'+/-1*!&&1**-*n:m~-~~*'*'if?~**~* -~~~*:~m:g*p,.~'4l""""'lP.
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~****
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- 1.::.
Proprietary information contained in separate submittal
- The two-phase local pressure drop, !.IPL, is calculated by:
K Az z
</JTPL Proprietary infonnation contained in separate submittal~
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- 5. 1
. QUESTION Provide comparisons of calculated parameters (e.g., relative power and reactivity
'coefficients) with experimental observations (critical facilities and reactor irradiations) for the core referred to in Section 5.2 of NED0-20103 which contain:
(1) mixtures of 6x6 and 7x7 bundles (2) mixtures of 8x8 and 9x9 bundles (3) only 7x7 bundles
RESPONSE
Comparison of calculated local parameters with measurements aregiven in Amendments 5 and 7 to the application for review of GESSAR.
- 5. 2 QUESTION Define the terms "controlled" and "uncontrolled" used in Figure*5.2 (,f NED0-20103.
Provide comparisons of multiplication factors as functions of void fraction similar to F.igure 5. 2 for the 50% controlled case.
RESPONSE
The values given in Figure 5.2 of NED0-20103 are for infinite arrays of the*
bundle types not.ad and as such controlled and uncontrolled means fully controlled and fully uncontrolled, respectively.
The k"' of a 4-bundle array with 50%
control fraction as a function of in-channel void fraction is given in the following table. The control blades inserted are the diagonally adjacent control b.lades.
2 *. 50 wt % U-235 Bundle, Infinite Lattice k at 50% Control Fraction (1)
Bundle Type 7x7 8x8 Infin:t.~e I.attice k <2>
0 20%
40%
- 1.012 0.997 0.977 1.026 1.014 0.992 NOTES:
(1). Diagonally adjacent control rods inserted 70%
0.937
- o. 956.
(2)
Basic array is 4*bundles of the type indicated whose exposure is 0 (3)
In channel void fraction
5.3 QUESTION Provide the assumptions and bases used to calculate the rnaximUJ!l local peaking as a function of exposure as shown in Figure 5.7 of NED0-20103, (e.g., was an infinite lattice of one bundle type assumed; what value of void fraction, what control rod program and gadolinia distribution were assumed?)o,Explain why the variatfon in maximum local peak with exposure is different for the two lattices.
RESPONSE
Single-bundle, two-dimensional irif inite lattice calculations were used to generate the data given in Figure 5.7 of NED0-20103.
In addition, an in-channel void fraction of 40% and the actual gadolinia distribution were used.
The bundle was depleted in the uncontrolled state to calculate the bundle isotopics at each exposure point.
The difference in behavior of the 2 lattices with regard to maximum local peak is caused. by the difference in position of the peak power tod.
As both bundle types are depleted, the position of peak 10cal power rods changes.
In the case of the 8x8 lattice the peak power r:J, excr.pt for high (> 11 GWd/t) exposure, is located away from the gadolinia rod.
For the 7x7 lattice, the peak power rod is located ~djaccnt to the gadolinia rod at exposures >8 GWd/t.
The early suppress.*Jn of the depletion in the vicinity of the gadolinia rod causes a higher relative power at high exposure.
Note that although relative peak local power increases in the 7x7 bundle, absolute power is decreasing due to the net decrease iri bundle reactivity and therefore power.
- 5. 4 - QUESTION What is the initial value of maxi~um local peaking and the var~ation of local pe~king with burnup in the initially u. axposed 8x8 assemblies placed in the array of exposed 7x7 assemblies? *compare this maximum lbcal peaking with that which would occur with initially unexposed 7x7 assemblies loaded in the exposed 7x7 array.
RESPONSE
Due to the watergaps between the fuel bundles, the effect of exposure mismatch (fresh 8x8 and exposed 7x7) on relative local peaking is small (maximum increase on the order of 1%).
Therefore, the relative local peaking is as given in Figure 5-7 of NED0-20103.
The response of a fresh 7x7 fuel bundle will be similar.
-- -~ *----
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5.5 QUESTION What is the maximum expected exposure of bundles which remain in the core, 2) t.ho the core, and 3) fresh 7x7 bundles which a 6-inch,axial segment of o*the 7x7 8x8 bundles which are to be loaded in could he loaded in lieu of 8x8 bundles.
If greater than 22 GWd/t, provide the maximum local peaking factor for each type of bundle out to its maximum exposure.
RESPONSE
The maximum exposures of a 6-inch segment of the various bundle types at the expected discharge states are given in the following table together with the relative local peaking values.
Maximum Exposure of a 6-Inch Segment of a Fuel Bundle at Discharge Bundle Type Exposure, GWd/t Power 7x7 (Initial) 26.2 7x7 (Reload-1).
Jl.3 7x7 (?.5 w/o) 33.6 (estimate) 8x8 (Reload-2) 33.6 Relative Local Peak
. at 35 GWd/t 1.16 at 30 GWd/t 1.10 1.15 (estimate)
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- 5. 6 QUESTION Provide the expected operating power level as a function of exposure of 1) th 1~
8x8 bundles and, 2) the 7x7 bundles which could have be,n used in lieu of 8x8 bundles.
,RESPONSE The power level of a fuel bundle.will depend not only on its exposure but also on location in the core, operating control rod pattern, etc. However, an average power level over an operating cycle can be predicted using Haling calculations and is given ill the following table.
Data for a 7x7 bundle will be similar to the 8x8 data below:
Maximum Corresponding Cycle of Bundle Average Average Exposure Range, GWd/t Operation Power, MWt BOC EOC lst 4.58 0
7.2 I
2nd 4.66 7.'l..
13.0 3rd 4.49 9.5 15.2 4th 4.21 15.2 20.s 5th 4.14
- 16. l 21.2 6th 2.96 21.2 25.1 7th 2.77 25.6 29.0
- 5. 7 QUESTION If. fuel shuffling is to be done, describe the procedures to be used.
What calculations are done to a) determine local and gross peaking factors, b) verify
., shutdown margin and, c) determine weighting factors used in calculating behavior following accidents involving significant spatial effects, such as a rod drop accident.
RESPONSE
At the time of the refueling, the fuel bundle types in each core location will be identical to that in the licensing reference loading pattern.
In addition, the average reactivity of the four fuel bundles surrounding each *control rod is maintained as close to the refe~ence loading patte_rn as possible.
When the final core configuration is known, control rod sequencing calculations are performed to determine and adjust the core power distribution.
In addition, shutdown margin calculations are performed and their accuracy confirmed at the site.
The methods employed by C-r1eral Electric do not involve the use* or weighting factors.
The cc*.* trol.rod drop accident analysis for example, employs the use of <)atial reactor kinetics and is described in detail in NED0-10527 "Rod Drop Accident Analysis for Large Boiling Water Reactors," Licensing Topic:::::. Report, March 1972.
Since spatial core kinetics are employed, th~ _weighting factors are an integral part of the calculations. If*fuel is. shuffled then the only requirement is *to show compliance with the bounding curves for *rod drop analysis as shown in "Dresden 3 Nuclear Power Station Second Reload License Submittal, dated September 1973.
QUESTIONS AND ANSWERS REFERRING To* ATTACHMENT c GESSAR QUESTIONS APPLIED TO DRESDEN 3 RELOAD 2 CONTENTS CONTENTS 3.69*
4.32 4.13 4.33*
4.14 4.34*
4.15*
4.35 4.16*
4.36*
4.17 4.37 4.18 4.38 4.19 4.39 4.20*
4.40 4.21*
4.41 4.22
. 4.42 4.23 4.43 4.24*
4.44 4.25 4.45 4.26 4.-+0 4.27 4.2R.
I'. 29 4.30 4.31
~-.;___,, ___... -..
- Not contained - to be submitted the week of December 31, 1973.
- 4. l?
QUESTION With respect to. Section 4.2.1.1.2.4 of GESSAR, justify the use of maximu~ shear stress theory and related stress intensity limits with respect to Zirconium fuel cladding.
Clarify whether the fuel cladding is considered a ductile material after exposure to reactor core operating conditions or is classed as a non~ductile material, with design, analysis and qualification procedures based rin brittle material behavior.
Section III of the ASME Code specifically excludes "tubes or other forms of sheathing used only for cladding nuclear fuel 11 and is not considered an applicable guide in this area.
RESPONSE
The subject stress criteria were adopted by GE during the mid 1960's and provide a very conservative design.basis.
Essential-ly all of the GE BWR fuel currently in operation was designed in com~lian~e with these criteria.
The resultant fuel rod designs have exhibited sat~sfactory dimensional stability, with no significant dimensional changes having been -0bserved.
~~garding the questiori of clad ductility, the clad has been demonstrated to be sufficiently ductile to accomodate the strains associat~d with the subject stress limits.
See NEDO 1 0 5 0 5,
11 Ex per i e n c e w i t h B WR Fu e l t h r o u g h S e pt ember l 9 71,
11 'w h i c h has been previously submitted to the USAEC.
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. 4. 14 QUEST I ON Provide fuel handling and shipping desig*n loads and their relation to stres~ and strain limits to establjsh that design.limits have not been violated during handling or shi~ping.
RESPOUSE Duririg shipping, the fuel bundle is in the horizontal position with flexible packing separators fitted between the fuel rods such that the weight of the rods is directly transmitted to the s~ipping container and is not transmitted through the fuel spacer grids.
The bundle shipping pro~edures were qualified by shipping tests performed on prototype fuel bundles representative of the Dresden 3 Reload fuel design.
- Each individual fuel bundle will be inspected at the reactor site to verify that no ~eviation~
of important dimensional ~haracteristics ha~e occurred as the result of shipping.
- ---*-~~
4.17.
QUESTION Provide the numerical ~alues used for the Zirc~loy cladd*ing yield strength and ulti-lnate tensile strength in conjunction with the stresi intensity limits.
In addition, state the cladding thermo-mechanical history and as~ociated temperature and fast ne~tron f)ux (or fluence) for which the stress limits apply.
RESPONSE
The cladding yield and ultimate strength correlations used*
.in desi~n are given by figures 4.17 A and B. The irradiated properties shown by these figures are applied for a cladding neutron fluence of 4 x 162 0 NVT (>l MEV), or greater, and for 0
a cladding temperature of less than 800 F, as noted.
Sin~e
- figures 4.17A and Bare GE proprietary they are submitted under separate cover.
4.18
- QUESTION List fuel rod deflection and cladding strain limits and provide justification for.their adequacy.
For example, 0.060 inches is given for the rod to rod clearance and a 1% cladding strain limit is ~given.
RESPONSE
The fuel rod-to-rod spacing limit of 0.060 is an internal G.E.
li~it based on the range of clearances that had been used in CHF testing (i.e. only clearances of 0.060" and above had been tested). More recent testing to clearances below 0.060 11 would indicate that a lower limit may be acceptable.
The justification for the.1% cladding strain limit is given in NED0-10505 "Experience with BWR Fuel Through September 1971 ~** which has been previously transmitted.
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- 4.19 QUESTION What is the strain limit in the axial or radial direction?
Since creep rupture strain depends on stress level, show how the stress effect is incorporated in the ~trairi li~it.
Total strain is not defined in the strain calculations.
If the strain limit was based on total strain, discuss how it is calculated.
RESPONSE,
I The definition and basis for the damage strain limit has been provided in N ED0-10505 11 Experience with BWR Fuel th rough September 1971.
11 Creep rupture has not been considered in the analysis because the events wherein the damage strain limit is applied are of short duration (i.e.., abnormal operational transients}.
- 4.20 QUESTION Is there a natural frequency limitation on the fuel assembly?. If so~ is it relatable to the. primary system frequency.
RESPONSE
There is no defined limit on the natural frequency of a fuel assembly.
The stiffness of all the fuel rods combined is only a small fraction of the stiffness of the fuel channel, ~o in seismic vibration analysis the fuel assembly stiffness is assumed to be that of the channel.
The weight of an 8x8 fuel b.undle is the same as the weight of an 7x7 bundle.
Since the assembly natural frequency is a function of weight and stiff-ness, the natural frequency of the 8x8 assembly is the same as that of an 7x7 assembly.
- 4. 22 1UESTION What is the deflection limitation on the channel?
RESPONSE
A report regarding channel deflections was recently submi.tted I
to the AEC by the Genl;!ral Electric Company (J. A. :Hinds, letter to D. J. Skovholt, "Interim Report - BWR Fuel Channel Defle_ctions, 11 November 28, 1973). This report provides the answer to this question.
,* i 4.23 QUESTION Give safety factors applied in the fatigue design, creep rupture, fatigtie creep interaction, and instability (buckling) analyses.
RESPONSE
- a) GE uses the fatigue design correl~tion developed by O'Do.nnel and Langer {"Fatigue Design Basis for Zircaloy Components," Nuclear Science and Engi'neering:
20, 1-12 (1964).
This correlation includes a safety.factor of 2 on stress, or ~O on cycles {whichever is more conservative).
Also, in the Dresden 3 R2 fuel design less than 5% of the allowable fatigue life is used.
b)
Creep rupture.and fatigue creep interacti-0ri have beeri considered in the design and have not been identified as important mechanisms.
The stress design limits used by GE provide margin from significant creep.
deformation of the fuel rod.
c)
The creep bucklin~ a~alysis for Dresden 3 R2 fuel has been provided in "Dresden Unit 3 Supplement C to the Second Reload License Submittal,"
(AEC Do~ket 50-249).
The maximum predicted axial compressive load in a fuel rod is a small fraction of the buckling load.
/
- ~.25 QUESTION Provide tables of material properties of both fuel rods and pellets as functions of temperature and irradiation. The pioperties should include modulus of el as ti city, Poisson's ratio,* ~henna l expansion coefficient, yield stress, ultimate stress, uniform ultimate strain, creep constants and creep equations.
RESPONSE
The cladding design strength is shown in Figures 4.17 A and B attached to the response to Question 4.17.
The creep equations and constants*
were provided in NEDM-10735 Supplement 1, 11Densification Considerations in BWR Fuel Design and Performance, 11 page 11-1. Other material prop-erties are shown in Table 4.25.
As previously noted, the GE fuel design has provided substantial creep resistance. Since creep has been deter-mined to be of no consequence in GE fuel it has not been considered in the standard design practice. Table 4.25 is GE proprietary and has been submitted under separate cover.
- 1.
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4.26 QUESTION Provide a more detailed description of the top guide and its conn.~ction with upper tie plate.
RESPONSE
The top guide is an "egg-crate" structure of stainless steel bars which form a 4-bundle cell.
The four fuel assemblies ar~
lowered into this cell and, when seated, springs mounted at the tops of the channels force *the channels intd the c-0rners
~f the cell such th~t the sides of the channels.~ontact the grid beams.
The upper tie plate is positioned within the channel by dimples pressed into the channel wall to center the fuel bundle, and restrict relative movement between the bundle and channel.
The attached Figure 4.2.6 shows a schematic of this arrangement.
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4.27 QUESTION Describe how the channel is connected to t.he fuel assembly.
What is the interferE!"n.ce dimension between the reusable fuel channel ~~d the fuel assembly and what are the expected inter~
ference loads?
RESPONSE
The channel is open at the bottom, and is installed by lowering dowri over the. fuel bundle.
At the top of the ~hannel, two diagorially opposite corners have welded tabs which support the weight of the channel from raised posts on the upper tie plate.
One of these ~aised posts has a threaded hole, and the channel is attached using a threaded channel fastener assembly, which also includes the fuel assembly positioning spring mentioned fo response
. to question 4.26.
The only interference that can exist between the channel ~nd fuel bun.dle is due to differential thermal expansion between the stainless steel upper and lower tie plates and the channel,.
and only in the case of adverse tolerance stackup.
This inter-ference has no effect on the reusability of the channel.
' 4. 28 QUESTION Provide the dimensions and s-pring constant of the fuel stack hold down spring.
RESPONSE
The following are design characteristics of th~ D3-R2 Retainer:
Spring Constant Free Length Wire Diameter Outside Diameter Numb.er of Turns
--~~-~-~.....
16 lbf/in 1 2. 04 II
- 062 II
. 396 11 33
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4.29 QUESTION Provide the specific materials use~ in the spacer g~id design.
. RESPONSE The spacer grid structure is Zircaloy-4 (ASTM B-352 grade RA-2).
T~e springs are Inconel X-750, which has the following nominal composition:
Ni + Co - 70% Min.
Co
- l % Max Cr 17%
Fe 9%
Ti *
- 2.25-2.75%
Nb + Ta - 0.7-1.2%
Al
- 0.4-1.0%
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4.30 CUESTI 011 Describe the grid lock on the water rod, and provide a drawing.
RE-SPONSE The spacers are spaced and retained axially by tabs which ar~ welded on the water rod.
The tabs are positioned between two structural members of the spacer to provide axial ret.ention.
(See Fig. 4.30).
The water rod has a square lower end plug to prevent rotation of the water rod and assure engagement of the tab.
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4.31 QUESTION The analytical calculation of fuel-clad mechanical interaction that is used to satisfy the design bases should be given. A detailed complete description is needed, including a general description, assumptions, mathematical equations, sequence of application of equations or a flow chart, sample calculation (bench mark) and a com-parison with test results.
RESPONSE
A discussion of the design models employed to predict the effects*
of the predominant thermal/mechanical mechanisms which result in clad loadings is presented in NED0-10505, "Experience with BWR Fuel Through September 1971."
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- 4. 32 QUESTION In addition to the fuel-clad mechanical interaction analysis, describe the calculations used to justify:
- 1.
Plenum volume and pressure calculation including plenum buckling stability and creepdown.
- 2.
local.stres~ of cladding at fuel pe~let interfaces, and relation to.observed cladding brittle failures.
- 3.
- claddin~ transient power stress-strain analysis.
- 4.
Axial ratcheting of fuel-cladding.
RESPONSE
- 1.
The available gas retention volume is deter~ined based on the following assum~tions:
- a.
Minimum allowable plenum length.
- b.
Maiimum expected fuel-clad differential axial expansion.
- c.
No annular volume between pellets and cladding.
- d.
Maximum velum~ displacement by components contained in the plenum (spring an~ getter).
The fi~si~n ga~ release mod2l was described in section 3~2.5 of "Dresden 3 N11clear Power Station Second Reload License Su b m i t t c l 11 T h e f u e l rod i n t er n a 1 pr e s s u re is ca l c u l a t e d using the perfect gas law.
Creepdown and creep collapse of the plenum are not considered beca~s~.significant creep in the plenum region is not expected.
The fuel rod is designed to be frea-standing.
The temperature and neutron flux in the plenum region are considerably lower than in th*e fueled region, thus the margin to creep collapse is substantially greater in the plenum.
Direct measurements of irradiated fuel rods have given no indication of significant creepdown of the plenum.
- 2.
An evaluation of local stresses at pellet interfaces is presented in NED0-10505 11Experience with BWR Fuel through September 1971, 11 May 1972.
- 3.
Transients up to 116% of rated power were evaluated by the bases in section 3.2.3 of "Dresden 3 Nuclear Power Station Second Reload License Submitted".
Abnormal operational transients exceedin~ 116% of rated power were evaluated in section 3.3.2 of the same document.
- 4.
Axial ratcheting of fuel cladding is not considered in BWR fuel rod design because it has not been observed.
Prototype fuel rods have been operated in the Halden te~t reactor with axial elongation transducers.
No significant axial ratcheting has been observed.
. 4. 35 QUESTION Discuss the behavior of the bottom fuel rod support during cyclic thennal axial expansion and contraction of the fuel rods.
RESPONSE
The loads transmitted to the lower.fuel support for cyclic thermal expansion and contraction of the fuel rods result in low stresses.
The lower fuel support is not limited by strength or fati~ue considerations for this type of loading~
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4.37 QUESTION Determine the thermal stress in the fuel rod caused by lateral thermal differential expansion between the tie plate. and the spacer grid.
RESPONSE
The thermal stress caused by lateral thermal.differential expansion between. the tie plate and the spacer ~rid is less than 400 psi.
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4.2i QUESTION Evaluate the -effects of fuel ro-d bowing together with spacer g~id response including time dependent behavior due to creep.
RESPONSE
For Dresden 3 R2 fuel, the maximum thermal bow stress* is 470 psi, the makimum thermal bow deflection is 0.0038 inch, and the maximum force exerted by the fue~ rod upon the spacer grid due to thermal bow is less than 0.1 pound.
This stress and force are too low to cause creep of cladding or spacer grid.
- 4. 39 QUESTION Explain in detail how the fuel rod flow induced vibration is not a problem in an 8x8 design.
Evaluate vibration of.::he shroud and other internals and their effects on fuel rod vibration.
RESPONSE
- The flow induced vibration design analysis is based on test data which encompasses the expected range of flow velocity and steam quality. The empirical correlations derived from these data predict a vibration amplitude of 0.7 mil.
The influence of other core components is not considered to ~ffect fuel rod vibration.
The test information and data are contained fo the following report: Quinn, E. P., "Vibration of Fuel Rods in Parall~l Flow, 11 GEAP-4059, July 1962.
f~~i 4.40 QUESTION Discuss the operating experience you have with the 8x8 fuel des~gn, including how mu~h experience you have with 95 TD uo2, 9 mil gaps, water rods, no dish pellets, and annealed Zr-2 cladding of 34 mil wall thickness.
RESPONSE
General flectric's operating experience with production and developmental BWR fuel is discussed in section 3.4 of NED0-20103 and summarized in tables 3-3 through 3-6 of NED0-20103.
Although there is no experience with production BWR fuel rods having the exact combi-nation of gap size, cladding thickness, and pellet diameter to be employed in the 8x8 fuel design, the experience obtained is for fuel rods of similar design which employ typically the same fuel ~nd claddi~g materials, have a range of design parameters encompassing those of ~he 8x8, and have been operat:~ *in typical BWR environments at combinations of power and exposure significantly exceeding those to be experienced by the 8x8 f_ue 1 design. Consequently, a 11 of the General El e~tri c BWR operating experience on zircaloy clad, high density uo2 pellet fue,. is considered to be applicable to fuel of the 8x8 design and demonstrates the propensity for high integrity fuel performance from this design.
In addition to the developmental fuel experience summarized in NED0-20103, there are ~ small number of developmental fuel rods typical of the 8x8 design currently undergoing irradiation in the Halden test reactor. These rods to date have successfully attained average expo:sures of 7000 MvlD/stu with maximum linear heat ratings of 14 to 16 Kw/ft.
Spacer capture water rods of a simi'l:ir design to that proposed for use in the 8x8 reload design have been employed in twenty-one reload fuel assemblies loaded in the Big Rock Point Reactor in early 1973 (reload 12).
These rods have performed satisfactorily to date.
In addition, General
4.40 Question (Continued)
Electric has subjected typical 8x8 water rods to out of pile flow testing at reactor operating conditions in ord~r to confirm hydraulic d~sigrr parameters.
-~.*
4
- 41 QUE Sf ION Describe in detail the experiences testing, and analyses, that provide an understandin~ of the failure mechanism of the fuel channel and measures taken to eliminate s~ch failures; RESPONSE.
It is assumed that this question is directed towards the recent G. E. channel experience in the KKM, Vermont Yankee, and Pilgrim reactors in which channel wear and cracking were observed. This experience has been attributed to a unique combination of design features employed only in these three BWR 1s and is, therefore, not applicable to fuel channels in other BWR's.
Infonnation and analysis concerning this channel experience to date have been submitted to the USAEC in the following references:
l *. J. A. Carroll of Boston Edison Co., letter to V. A. Moore of th~
USAEC,
Subject:
Nov. 14, 1973 Docket #50-293.
- 2.
D. E. Vendenburgh of ~ermont Yankee Nuclear Power Corp., letter to USAEC Directorate of Licensing,
Subject:
Vermont Yankee Channel Wear, Investigation and Corrective Actions Taken.
4.42 QUESTION Provide the percentage gadolinia per pellet, the number of gadolinia-poisoned rods per assembly, number 9f assemblies containing poisoned rods, total number of assemblies, axial distribution of the gadolinia in the rods, location of the gadolinia-poisoned rods, and operating limits on the poisoned rods, i.e., linear powers at various stages and powers calculated to result in* incipient melting.
RESPONSE
Information concerning th~ number of gadolinia-poi~oned rods per assembly, number of assemblies containing poisoned rods, total number of assembl.ies and the location of the gadolinia-poisoned rods was provided in the following two submittals:
- 1.
Dresden Unit 3 Second Reload Submittal, AEC Docket 50-249,
- September 14, 1973.
2..
Dresden Unit 3 Supp 1 emen t A to Second Re 1 oad License Su bmi tta 1,
AEC Docket 50-249, December 6, 1973.
GE proprietary i nfonnation in this section of the response is provided under separate cover.
, Values for the limiting linear powers calculated to result in incipient fuel melting and 1% diametral cladding plastic strain are given for Gd203-uo2 fuel rods in Table 4.42-1.
I I,
TABLE 4.42-1 LIMITING LHGR FOR GADOLINIA-URANIA FUEL {KW/ft}
- 1 % Plastic Exposure Incipient Strain (MWd/STU}
Center-Melting of Cladding 0
- 18. 5
~ 22.5 20,000 18.0
~ 21.0 40,000 17.0
~ 18.0
- ~---*
4.43 QUESTION.
Discuss th~ operating experience of BWR 1s with gadolinia~
poisoned fuel, including reactor names, loading dates, fuel burnups
- and maximum powe~s.
RESPONSE
Production fuel rods e~ploying gadolinia-urania fuel pellets have been in use since 1965.
During this time~ a substantial number of gadolinia-urania rods have been successfully irradiated to appreciable exposures. Table 4.43 summarizes this experience.
Of these irradiated gadolinia-urania rods, only a sm~ll humber have experienced failure, none of which could be attributed to the fact that they contained gadolinia bearing fuel p~llets. The frequency of gadolinia-urania fuel rods experiencing failure has been typical of the statistically small failu: e frequencies experienced in uo2 fuel rods, 'V(),2% *
,\\
- i
- ~:-.::."-.?.: *..:.:..:;:.:::...:..:.
.:::.:::~:.~::
~:?:?_;-;:"I:\\G ~-~,::;\\.:;:::::: ;.;:~ PRCI:'CC'i'ICS. G..;DJLI~:IA-BL\\?,I~:G Fl~L AS OF ~~;~c;.: 31, 1973 Hi.;::!:O 1 ct f';.;e 1
. r:.. :e Il!f(a) v(a)
EG(c)
F 11 I
- Tsur.;c;a J~
J;.a Fuk:ishi~a-1 TXA Dresce~-2 CY D'I
~tir.e ~ile Pt.
~:*~
GEA Oyster Cr-;*k Jc;. **
tiucle~cr GEA Mfllstc~e GEA Gt:3 Quad Cities CX Quac Cities 2 CY F:..:e1
~:;~
r::a:-e~e..
(iri:~es) c:""c
- ..,,... ~.J c:.;c:
.5525
.5525
.563
.553
.553
.563
.563
.553
.553
.563
.563
.563
- 553
.553
.553
.* 553 (a)
Ex~¢sures to 1/1/72 Ci~~;ir.g ir.:c:.::"ess
( ::-: ; ; s l 35 35 40 40 32 32 32 32 32 32 32 32 32
. 32 32 37 32 32 Fe::e:-to f..c~: ":e Fi.;el Cla~=;~; Gap
.Le~gth
(~ils)
(ir.ches) 10 10 11 11 11 12 12 12 12 12 12 12 12 12 12 12
. 12 1°2 108;25 108.25 70 70 79 144 144 144 144 144 144 144 144 144 144 144 Total Su!':'!:er of Roes 3,7.. 4 3,816 2,926 3,696 6,336 2,352 2, 156 784 10.535 24,9-11 1, 175 1,950 7,5.;.;
- l. 372 4,018 1,470 35,476 35,476 1:1.;-:--ber cf GaG.:llinia-Bearino Rods 104(~)
324 lSO(c) 192 352 192 176 64 442 15.2. 7 96 120 624 84 246 90 1,760
- 1. 760 Tir.e in Core
- (years) 7.0 5.0 3.4 1.5 4.0 1.5
.5 1.2
- 1. 7 1.0 1.3
.5
- 1. 3
.6
- 1
- 1
.9
.9 Ex:osure r.!xl~~
~~xi~:.:~ ;..ssy/.E~!t Fl;,;x f..verac~ Assy.
(Oesic;~)
(:-~:c!Te)
(Sti;!h-f~2) 259~0/195'.:0 196CC/13250 151:0/l 0100 11900/6600 1 s10011 ~oco 3S'J.~0'J
- i:o.c::o 410,CCO 410,CCO 3e9,OQO 9650/8650 *
~03,c:o 2850/2350
~05,000 4950/2100
~CS,000 6350/6150 405,0CO 4300/3S50 405,COO 6750/65:0
~05,000 520C/~350 405,080 65C0/ 5So 405,oo*a 3950/2550 405,0CO 25J/200 405,000 250/200
- 405,COO 8350/3600 405,000 1850/1650 405,000 Pe:!:.
LP.S~
(Cesic;n)
{:<.Uft) 15.S 15.5
- 17. 7
- 17. 7
. 16:8 i 7. 5.
- 17. 5 17.5
- 17. 5
- 17. 5
- 17. 5
- 17. 5.
- 17. 5
- 17. s 17.5
- 17. 5 17.5.
17.5
{b) lncl:;c!es 98 stan~ard asserr!>11es wfth 1 Gd2o3-Alumfna rod; 2 special assemb11es w1th 1 Gd2o3-Alum1na segmented rod; and 4 special asse::tiltes with 1 Gd2o3-t:rania se~ented rod.
(c)
Inch:ces 35 standard asser.!>lfes with 4 Gadolfnfa-~z rods, 4 special MEG assemblies with 4 Gado11nfa-L'Oz.. rods, and 3 special EG asser.:!>lies with 8 ~. "o, 'n1'
- U
c **
t,J'"'"'
- 1
""'- vz. r. a...
(d)
Ex~osures to* 3/1/72
- ~
- 1
,j
~
~1 i
I i
I 1 -
[
.;.r----
4.44 QUESTION.
Discuss the results of all post-irradiation examinations of irradiated gadolinia-poisoned rods to date including visual examinations, neutron radiography, dimensional measurements, fission gas analysis, ceramographic examinations, radial grind isotopic sampling, and microprobe analysis.
RESPONSE
The results of all post-irradiation examinations of irradiated U02-Gd203 fuel rods examined_to date indicate no significant difference between the performance of the low concentration. (<5 w/o) Gd2o3-uo2 fuel rods employed in G.E. BWR's and comparable U02 fuel rods.
Details of these examinations have been )reviously submitted to the AEC in response to question 38 of NE0~-20135, "Responses to AEC Questions NEDM-10735 Supplement 1.
11
~~l ~-,.." ~ *. : *.' ::~_t :~*~*'-.*~:t.~
~ ~~~,\\; *.
4.45 QUESTION Thermophysical properties for the spe~ific Gdo1.5-uo2 solid solution compositio~ to be used should be presented and compared to the pure U02 propertiei. Properties of interest include melting behavior, thermal expansion, specific heat, vapor pressure and creep behavior.
RESPONSE
The concentrations of Gd2o3 to be employed in BWR fuel are small ( <5 w/o).
For these low concentrations of Gd2o3, the only thermophysical properties of Gd2o3-uo2 fuel considered to be signi-ficantly different from those of uo2 fuel are the thermal conductivity and melting temperature..The va 1 ues for these properties employed in design are compared to U02 in Table 4.45-1, which is GE proprietary and submitted under separate cover.
,