ML19329A868
| ML19329A868 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/21/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329A865 | List: |
| References | |
| NUDOCS 8001150761 | |
| Download: ML19329A868 (28) | |
Text
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IrlTERIM SAFETY EVALUATION REPORT ON EFFECTS OF FUEL ROD BJWING ON THERMAL MARGIN CALCUL?TIONS FOR LIGHT WATER REACTORS e
O 8001150 76[
CONTENTS 1.0 Introduction 2.0 DNSR Reduction Due to Rod Sow 3.0 Application To Plants In,The Construction Permit And Operating License Review Stage 4.0 Application To Operating Reactors 5.0 References i
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1.0 Introduction Data have recently been presanted to tne staf *~ wnicn srow that previously developed methocs for accounting for tna effect of fuel rod bowing on departure from nucleate boilino io a vssurizea e
water reactor (PWR) may not contain adequate thermal margin when unheated rods are present (such as instrument tubes).
Furtnce experimental verification of tnese data is in progress.
However an interim measure is r,,equired pending a final decision on tne validity of tnese new data.
The staff has evaluated tne impact cf these data ur, tne l
performance of all operating pressurized aater reactor Fode s for treating the effects of fuel rod bowing on nermal-nydraulic performance have been cerived for all operating 2'..'Fs.
These T.ccess are based on the propensity of the individual fuel designs tc bo.: and on the thermal analysis methods used to predict the coolant conditions for both normal operation and anticipated transients.
As a result of these evaluations the staff nat concluded that in scre cases sufficient thermal margin does not now exist In these cases, additional thermal margin will be required to assure, "rita nicn confidence, that departure from nucle 3ta bc linn ';,';3 ) Joes no t occur during anticicated transients.
This recort dist.sses "cw nuse conclusions were reacned and identifies tne arcont nf acci noril marcin required.
The mcdels and the recuired DNSR reductions wnich rest.'t from these models are meant to be only an interim measure until more data are available.
Because the data base is ratner sparse, an attempt was made to treat this proolem in a conservative way.
The required DNBR reductions will be revised as more cata become available.
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, The staff review of the amoL..t and consequences of fuel rrd bowing in a boiling water reactor is nas underway.
At present nn conclusions have been reached. When this review reaches a stage where either an interim or final conclusion can be reached, the results of this review will be published in a separate safety evaluation report.
It shou d be noted that tnroughout the remaii der of tnis report, all discussion and conclusions apply only to pressurized water reactors.
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_3.
2.0 DNBR Reduction Due To Rod Bow
2.1 Background
In 1973 Westinghouse Electric presented to the staff the results of experiments in which a 4x4 bundle of electrically heated fuel rods was tested to determina the effect of fuel rnd bowing to cnntact on the thermal margin (D BR reduction) (Reference 1).
The tests.vera done at conditions representative of P'AR coolant conditions.
The results of these experiments showec that', for the highes t powar density at the highest coolant pressure expected in a Westinghouse rea : tor.the DNBR reduction due to heated rocs bcwed to contact was approximately 8%
Fuel bundle coolant mixing ard heat transfer comouter procrams such as CDBRA IIIC and THINC-IV were able to predict the rasults of these exoeriments.
Because the end point could be predicted, 1
i.e.,
the DNBR reduction at ccntact,there was confidenca that tnp DNBR reduction due to partial bow, that is. bow to less than contact could also be corre,ctly predicted.
On August 9,1976 Westinghouse met witn the staff to discuss further experiments with the same. configuration of fuel bundle (ax4 )
using electrically heated rods.
However, for this set nf axseriments one of'the center 4 fuel rods was replaced by an unheated tuce of the same size as a Westinghouse thimble tube.
Inis new test confiauration was tested over the same range of power. flow and crassure as tne earlier tests.
However, with the unheated, larger diametsr rod the reduction in DNSR was much larger than in the earlier (1973) tests.
- 4 The data consisted of points corresponding to no intentional bowing (that is, a certain amount of bowing due to tolerances cannot be prevented) and to contact.
No data were taken at partial clearance reductions between rods.
On August 19, 1976 CE presented results of similar experiments 5
to the staff.
These tests were performed using a 21 roc bunale of electrically heated rods and an unheated guide tube.
Results were presented for not only the case of full contact, but also the case of partial bowing.
The staff attempted to calculate the Westingrouse results with the COBRA IIIC computer code but could not obtain agree-ent with the new data. Westinghouse was also unable to obtain agreerent between their experimental results and the THINCIV comouter coce.
Both sets of data (Westinghouse and CE) showed similar effects due to variations in coolant conditions.
For both cases, the DNSR reduction became greater as the coolant pressure and the rod cower increased.
Because both sets of data showed that plant thermal margins might be less than tnose intended, the staff derived an interim model to conservatively predict the ONBR reduction.
Since the data with unheated rods could not be predicted by existing analytical methods, empirical models were derived. These models give the reduction in DNBR as a function of the clearance reduction oetween adjacent fuel rods. Two such models were derived, one based on the Westinghouse data and one based on the CE data.
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. 2.2 Model Based cn Westinahouse Data Data were presented by 'lestinchause for the DNBR reduction at full contact and with no bow.
No data at partial gao closure were presentec. Westinghouse proposed, and the staff accepted, a straight line interpolation between nese two coints as snown in Figure 2.1.
This approach is conservative since one would expect One actual benav-ior to more r.early folicw a curvec line as sect;n in the same fi:c.rn.
The DNBR reduction would increase slowly in macnitude as tne fuel rocs bowed to contact. As the rods cecome close er.oucn so tnat there would be an interaction between the two rods, the CNER reduction woulc tnen increase more rapidly.
No pnysical mechanism nas been postulated which would lead to sudden large cecreases in the DNB? for smali or moderate gap closures.
Thus, tne straicht line approximation is believed to be an overestimate of the expected behavice All manufacturers of reactor cores, including Westinanouse, include a factor in their initial core design to account for the reduction in DNBR tnat may result fron citen reduction fro.-
fabrication tolerances and initial rod bow.
The amount of tnis pitch reduction factor varies with the fuel design and the analysis methods which are used.
For any particular core tnis facter is not varied as a function of turnuo.
s 6-In developing the interim rod bow penalties described in this report, it became apparent tnat the penalty should be a function of burnup since tne magnituq,e of rod bow is a function of burnuo.
However, to maintain existing thermal margins early in core life wnen only a small amount of fuel rod bow is anticipated, tne initial pitch reduction factor was incluced until sucn time as t*1e ros bow DNBR reduction became greater.
This is representec as tne stra1cnt horizontal line on Figure 2.1.
2.3 Combastion Enoineerino 'tudel Combustion Enginee-ing performed experiments to determine the effect of rod bowing oc DNBR wnich included some cases in which the effect of partial bowing as well as bowirc ;o contact was determined.
Again, a straicnt line interociation is usec.
However, the point of zero DNBR reduction is not at zero clearance reduction but rather, at an intermediate value of clearance reduction. This is shown schematically in Ficure 2.2.
The horizontal straight line, representing the initial pitcn reduction factor is included as explained previously (Section 2.21.
. 2.4 Models for Cabcock and Wilcox and Exxcn On Aur ust 17,1C73 renrascatc tivat of WcocF an<: ' fi l c o;-
met with the staff to discuss this problem.
Babcock and Wilcox did not present any data on the effects of rod bowing on DN5R.
Iney nad previously presented data to the staff on the ancunt of bowinc to be expected in Babcock and Wilcox 15x15 fuel assemolies.
Because Babcock ard Wilcox had no data on the effect of rod tow nn DNSR.
the staff applied the Westinghouse model to calculate the effect of rod bowing on DNSR 'or Sarcock and Wilccx fuel.
The amount of fuel rod bowing was calculated using :ne Saccack ano Wilcox 15x15 fuel bundle data.
Representatives of the Exxon Nuclear Corporation discussed the effects of fuel rod bowing in tne presence of an unheated red.cn DNBR with the staff en August 19. 1976. Exxon nas no data pertin-nt to this problem.
Exxon has not performed DNS tests witn owec rocs from H.
3.
The first cycle of Exxon fuel nas just ceen recoved macnitude of rod Robinson and the results of neasurem=nts on int-The effects bowing have not yet been presented to the staf f.
of fuel rod bowind for Exxon fuel were avaluated on a plant by plant basis as discussed in Section a.O.
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- 2. 5 Apolication of the Rod Bow /DNBR 5'odel Using these empirical models, the staff derivea ONBF.
reductions to be applied to both cperating reactors and plants in the Construction Permit and Operating License review staae.
The procedure in applying these empirical models is as follows:
Steo 1.
Predict the c'learance reduction due *o rod boa as a function of burnup. An expression of the form 2+$=a+b7[3U
'o is used where
- C fractional clearance recaction due to red bewina
=
Co empirical constants obtained for a civen fuel cesicn a,b
=
BU = burnup (region average or bundle averace. docencinc on the fuel desigrer).
Steo 2.
Apply the previously discussed emoirical mouels of DNSR reduction as a function of clearance reduction usina tne value of.iC/C calculated from step 1.
g Step 3.
The staff has permitted the reduction in DNER c=lculate1 in step 2 to be offset by certain available thermal narqins.
These may be either generic to a civen fuel desian or clan decendent An example of'a generic thermal margin whic aculd be usod
- .c offset the DNSR reduction due to rod acw is tne fac* that the DN30 limit of 1.3 is usually areater *han the value of DNSR acove wnicn 95% of the data lie with a 95' confidence.
The differonce between 1.3 and this number may be used to o## set the ONBR raduction.
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An example of a plant specific thermal margin would be core flow greater than the value given in the plant Technical Specifications.
A discussion of the application of this method to Construction Permit and Operating License reviews is given in Section 3.0.
A discussion of the application and the results of this metnod to operating reactors is given in Section 4.0.
The application to reactors using Exxon fuel is also discus' sed in Section c.0.
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_ 10 3.0 Application to Plants In Construction Permit And Goerating License Review Stage 3.1 CP Applications No interim rod bow DNB penalties should be applied to CP applications. The rod bow data upon which the interin limits have been based should be considered prelimirary.
There is sufficient time available to review the data and assess a penalty, if any, prior to the OL stage. We will advise each CP applicant of the nature of interim penalties being applied to OL reviews and aparating reactors.
Since it appears that power derating is not necessary, there is no need to require design commitments at the CP stage; however, since limitations on operating flexibility may be required, we will need commitments from the applicant to (1) fully define tne gap closure rate for prototypical bundles, (2) determine by experiments the DNS effect that bounds the gap closure from part (1), anc (3) apoly any calculated loss of tnertal margin from steps (1) and (2) to reactor transient analyses. Such commitments should be part of our CP review effort.
3.2 OL Acolications Plants which are in the operating license review stage snould consider a rod bow penalty.
This penalty should be as cescribed in Section 2.2 for Westinghouse or Section 2.3 for Combustion Engineering. Babcock and Wilcox piants should use the rod bow vs.
burnup curve appropriate to their fuel and tne Westincncuse curve of DNBR reduction as a function of rod bow.
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All applicants may propose arcropriate tnernal margins (as discussed in Section 2.4) to help offset the calculated DNBR reduction. DNBR reductions could be greater fcr plants in the OL review stage than for a similar operating plant because plant specific thermal margins cannot be used to help Of fset the ONBR reduction resuYting from application of the mod 91.
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. 4.0 Acolication To Operating Reactors The section divides the operating plants intn distinct categories and lists them according to the fuel manufacturer or reactor type. Operating plants which cannot be 50 categnrizet (such as plants with fuel supplied by more than one vendne) are placed in a separate category. T;neplantsassignedtoeachcategoryare listed in the appropriate subsection.
The conclusions reacned in this section are in sonw cases deper. dent on conditions c - analyst wnicn are valic nnly for ne present fuel cycle. Hence, the FtH or DNER reductinns wnich are given (or the fact that no such reduction is concludac to bo required) is valid only for the present operating cycle.
4.1 Westinghouse LOPAR Fuel The designation LOPAR stands for low parasitic and refers to the fact that the guide tubes in the fuel buncle are made of Zircaloy.
Table 4.1 gives a list of the operating plants which fall into this classification.
TABLE 4.1:
PLANTS WHICH CURRENTLY USE THE WESTINGHOUSE LCDAR FUEL ASSEMBLY 15x15 17x17 D. C. Cook Cycle 1 Trojan Cycle i Zion 1 Cycle 2 Seaver '/ alley Cycle i Zion 2 Cycle 1 Indian Point 3 Cycle.I Turkey Point 3 Cycle a Prairie Island 2 Cycle 2 Indian Point 3 Cycle 1 i
. TABLE 4.1 (cont.)
15x15 Turkey Point a Cycle 3 Surry 1 Cycle a Surry 2 Cycle 3 Kewaunee C'ycle 2 Point Beach 1 Cycle 5 Point Beach 2 Cycle 3 Prairie Island 1 Cycle 2 The reduction in DNSR due to fuel rod bowirn is assu ec to vary m
linearly with the reduction in cleararce te: ween :ne fuei rods (or fuel rod and thimble rod) according tc tne model discussed in Section 2.2.
The maximum value of DNSR reduction (at centact), obtained from the experinental data was used to calculate the DNBR recuction vs. bow for the 15x15 LOPAR fuel. This DNBR contact reduction was adjusted for the icwer heat flux in the 17x17 LOPAR fuel.
The clearance reduction is conservatively assumeo to ce given by the following equation for the 15x15 (and 14x14) fuel.
hh =
a+bIE5I wnere ;[
is t;.e ~ reducticn in clearance Co Su is the region average burnuo and a,b are empirical constrants fitted to Westinannuse 15x15 rod bow data
r 14 -
For the 17x17 LOPAR fuel, tne clearance reduction was calculated from the equation:
I (CD) 15x15( T }15x15 SC j
L X
LC/Co =
{)
t 17x17 where L = the distance between grids I = moment of inertia of fuel rod s hnv.aq the craft new On December 2, 1974. Westinanouse infnemally data pertaining to tne magnitude of roc tow as a functirn or req 1on avarage burnup in 17x17 fuel assemblies.
Th s cata sno... :nat tne aoove correction is probably conservat9/e and tnat ina 'nacnitada n-fuel rod bowing in 17x17 fuel roos can cettee na rocrey-ntec nv nn empirical function.
Tnis review is now uncerway The calculated DNBR reduction is partially offset by exist nr thernal margins in the core design.
For the Wes tincnouse L'P A' im-design scme or all of the folicwinq items were used in calculatina the thernal rargin for the cperating plants:
design pitch reduction conservatively chosen TCC used in desicn' Critical heat flux correlation sta*.istics (assured i-tner al analysis safety calculatiens) are more conservative tr.an required.
Densification pcwer spike facto,- included althoucn n ICnce' required After taking these factors into account, tne reductions n F1.1 i
shown in Taoie a.2 were found necessary All oceratinc plants listed in Table a.1 will be required to incorporate these reducticns in FsH into their present operatino linits.
- TDC (tnermal diffusion coefficient) is a me3oure.cf tne amcunt of mixina between adiacent subcnannels.
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TABLE 4.2:
FiH REOUCTION FOR WEST ~.NGHCUSE LOPAR FUEL CYCLE REOUCTICN IN FaH (1) 15x15 17x17 ZION 1A2 _
lst Cycle (0-15 Gwd*/MTU) 0-2 ramp 1-13 ramp 0-6 ramo 2nd Cycle
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15 8
(15-24 Gwd*/MTU) 3rd Cycle (24-33 Gwd*/MTU) 6 15 10 These reductions in F P may be treated on a recion by recion basis.
If tne licensee chooses, credi may be taken for the narain between the actual reactor coolant flow rate and the flow rate used in safety calculations. Credit may also be taken fcr a difference between the actual core coolant inlet temperature anc that assured in safety analyses.
In taking credit for coolant flow or inlet temperature margin, the associated uncertainties in these cuantities must be taken into account.
4.2 Westinchouse HIPAR and Stainless Steel Clad Fuel The designation HIPAR stands for hich parasitic and refers to tne fact that the guide tubes in the fuel bundle are made nf stainless steel.
These two fuel types, HIPAR and Stainless Steel clac. are grouped tocetner because the amount of bowing exoected (and observed) is significantly less than that in the observed Westinghouse LOPAR fuel.
The plants which fall under this classification are listed in Table 4.3.
- Gwd Pwd
- 1000 MTU MiU
-1 6-TABLE'4.3: HIPAR AND STAINLESS STEEL PLANTS Ginna Indian Point 2 San Onofre Connecticut Yankee The model for the reduction in DNB7 due to fuel rod bowing is assumed to be identical to that used for the LOPAR fuel.
For reactors in this category, the peak recuction in CN3R (correspondina to 100" closure) was adjusted to correspond to tre peak overpower heat flux of that particular reactor.
The amount of rod bowinq for the plants listed in Table a.3 wnicn use HIPAR and stainless steel fuel, was calculated by means of an adjustment to the 15x15 LOPAR formula.
This adjustment took the form of the ratio amount of bow for assembly tyDe
= (L/IE) assy type amount of bow for LOPAR fuel (L/IE)
LnpAR where L is the span lengtn between grids I is the moment of inertia of the fuel rod E is the modulus of elasticity of tne fuel rod cladding Ginna Cycle 6 The Ginna plant is fueled with 121 fuel assemblies.
Two of these are Exxon assemblies', and two are B&W assemblies.
The remainder are Westinghouse HIPAR fuel assemblies.
The experimental value of CNER reduction was adjusted for heat flux and pressure from peak experimental to actual plant conditions.
Ginna took credit for the thermal margins due to pitch reduction, desian vs. analysis
values of TDC and fuel densification po..er spike.
These thermal margins offset tne calculated CNBR recuttien so that no reduction in FaH is required.
San Onofre Cycle 5 San Onofre is fueled witn 157 buncles of 15x15 stainless steel clad fuel.
The experimental value of DNBR reduction was adjusted fnr heat flux and pressure from experimental to actual plant conditions.
San Onofre took crecit for the tnerma' rarcins due to oitch recuction and tne fact that a value of 1.75 was usec for F?.H in tne safety analysis while a value of 1.55 was usec in tne Technical Scecifications.
Because of adequate thermal margin, no reduction in F h is required for San Onofre.
Indian Point 2 Cycle 2 Indian Point 2 is fueled witn HIO'R fuel butidles.
The experimental value of DN3R recuction.;as adjustes for neat flux anc r
. pressure to actual plant conditions.
Indian Point finit 2 had thermal margin to offset this ONBR reduction in pitch reduction, design vs. analysis values of TCC, fuel densification power spike and a value of FAH of 1.65 useo in the desicn (vs. 1.55 in the Tech Spec). Therefore, no reduction of F2H is required for Indian Point Unit 2.
Connecticut Yankee Cycle 7 Connecticut Yankee is fueled with 157 stainless steel clad fuel assemblies.
The DNSR reduction at contact was assured to be that used for the Westinghouse LOPAR 15x15 fuel.
No adjustrent..as made for heat flux.
The value of pressure was adjusted to the overpressure trio set point value of 2300 osi.
Full closure will not occur in stainless steel fuel out to tne design cuenup.
Connecticut Yankee has sufficient thermal margin in variabie overpressure and overpower trip set points to accommodate tne calculated DNBR reduction. Therefore no penalty is re"uirec.
4.3 Babcock and Wilcox 15x15 The reactors listed in Table 2.4 are fueled witn StW f ael.
TA3LE 4.4:
REACTOR USING B&W FUEL Oconee 1 Cycle 3 Oconee 2 Cycle 2 Oconee 3 Cycle 1 Rancho Seco Cycle 1 Three Mile Island '
Cycla 2 Arkansas 1 Cycle i
. The staff has reviewed the extent of rod bcwing whicn occurs wi th B&W fuel.
Based on this review, t.he following equation was derived for tne clearance reducticn be. ween fuel rods due to fuel rod bowing as a function of burnup:
= a + b '\\[ Bu Co where 1C is the fractional amount of closure Co Su is the buncle average burnup and c b are empirical constants fitted to 35.. cata The recuction in DNBR cue to fuel rod bowing is assumec to v.2ry linearly with the reduction in clearance between the fuel rces 'or fuel rod and thimble rod) but can never be lower than tnat cue o tre v tcn reduction factor used in thermal analysis, as explained in Secticn 2.2.
Babcock and Wilcox claimed and the staff acprovec crecit for the followina thermal margins:
. Flow Area (Pitch) recuction
. Available Vent Valve credit
. Censification Power Soike removal
. Excess Flow over that used in safety ar.alyses
. Hicher than licensed power used for piant safety analyses Based on this review and the tnernal margins cresented Dy 5&W to of fset the new ')estinchouse data, Rancho Seco is tne only Diant for which a reduction in DNBR is recuired. Table 5 c ves the values for the reduction of DNBR required at this time.
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TABLE 5:
DNBR REDUCTIONS FOR P&W Pla:;TS Burnup DNBR Reduction R a ncro S e_c_o.
Gwd Cycle 1 (0-15 WJ )
0 Gac Cycle 2 (15-24 MTu }-
1.6 Cycle 3 (24-33 Gwd )
3~
379 Rians must be sutritted to tne staf# to establisn how tnese reduction in 0 fBR will be accom ocatsc.
4.4 Comoustion Engineerin laxl Combustien Engineerinc nas presented cata to tne staf f on tne amount of rod bowing as a function of burnup.
The staff used tnis data to derive the following model for CE 12x1; 'uel.
AC a'
b -/ 5 u,'
Co 3C/Co = fraction of closure for CE fuel Su is the bundle average burnup and a,b are emoirical constants fitted to CE dat3 CE ras given credit #ce tnertal marcin cue to a multi,11er s' l.C65 on tne hot cnannel entnaloy rise used to account for citch reduction due to manufacturing tolerances.
Table J.6 presents tha required reducti~cn in DNBR using tne ccel describec loove. after accounting for this thermal margin.
Taoie 4.7 is a list of the reactors to which it applies.
A licensee planning to coerate at a burnup areator tnan 2400')
Mwd /MTU should present to the staff an acceptable method of
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accommodating the thermal margin reduction shom in Table 4.6.
This may be done as part of tne reload submittal if this burr.sp will not be obtained during the current cycle.
TABLE 4.6:
EFFECT OF ROD BCWING ON DNBR IN REACTORS WITH C0":USTION ENGINEERING 14x14 FUEL BURNUP REDUCTICN IN DN5R Cycle 1(0-E5kyh) 0 Cycle 2(15-24$.ou)
EA 0
G*d 3'
Cycle 3 (24-33 Tu, )
TABLE 4.7:
PLANTS FUELED BY CE FUEL TO WHICH VALUES OF TABLE 4.6 APPLY St. Lucie 1 Cycle 1 Ft. Calhoun Cycie 3 Millstone 2 Cycle 2 Maine Yankee Cycle 2 Calvert Cliffs 1 Cycle 1 4.5 Plants Fueled Partially With Exxon Fuel Palisades, H. B. Robinson, Yankee Rowe and D. C. Cook are cartially fueled with Exxon fuel. K discussion of these reactors follows:
Palisades Cycle 2 The Palisades reactor for Cycie 2 is fueled witn 136 Exxon fuel assemblies and 68 Ccmbustion Engineering fuel assemblies.
The Combustion Engineering fuel was treated accordinc to the Combustion Engineering model for both extent of rod bow as a function of burnup and CNBR reduction due to clearance reduction.
m
. The Exxon fuel was assumed to bow to the same extent as the Combustion Engineering fuel. This assumption is acceptable since the Exxon fuel has a thicker cladding and other design features which should render tne amcunt of bowing no greater than in the Combustion Engineering fuel, The DNBR reductti.n was assumed to be linear with clearance reduct. ion according to the Westinghcuse type curve of Figure 2,1.
The DNBR reduction at contact was based on the Westir.ghouse experimental data adjusted for the peak rod average heat flux in Palisaces and for the coolant pressure in Palisades.
The overpressure trip set point in Palisades is set at 1950 psis At this pressure the magnitude of the required DNBR reduction is greatly reduced.
The limiting anticipated transient in the Palisades reactor results in a DNBR of 1.36.
The thermal margin between this value and the ONBR limit of i.3 results in adequate thermal margin to offset the rod bow penalty.
Yankee Rowe Cycle 12 Yankee Rowe is fueled with 40 Exxon fuel assemblies anc 36 Gulf United Nuclear Corporation fuel assemblies, The fuel assemblies consist of 16x16 Zircaloy clad fuei rods.
The reduction in DNBR due tn fuel red bewing was assumed to vary linearly with the reduction in clearance between fuel rods The peak s
experimental conditions used in the '.Jestingnouse test were used to
- ~x the penalty at full closure.
The calculated reduction in DNSR is still less than that which would produce a DNBR less than 1,3 for l
l I
the most limiting anticipated transient (two pump out of four pump loss-of-flow),
Thus, no penalty is required.
H. B. Robinson _
Cycle 5 H. B. Robinson is fueled with 105 Westinghouse fuel assemblies and 52 Exxon Nuclear Corporation fuel assemblies, Tne Westinghouse 15x15 DNBR penalty model was applied to tne Mestinghouse fuel with a correction for the actual heat flux rather than the peak experimental val ues.
The Exxon fuel was considered to bow to the same extent as the Westinghouse 15x15 fuel so that the Westinghouse bow vs. burnuo equation was also applied to the Exxon fuel.
Inis assumption is conservative since tne Exxon fuel has a thicker cladding and other design features which should render the amount of bowin; no greater than in the Westinghouse fuel.
The DNBR reduction calculated by this method was offset by the fact that the worst anticipated transient for H. B. Robinson results in a DNSR of 1.68.
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. 5.0 References 5.1 Hill, K. W., et. al, "Effect of a Bewed Rod on DNB", Westinghouse Electric Corporation, WCAP 8176.
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' FIGURE 2.1 WESTINGHOUSE MODEL 0 DNBR
=
m Z
W CURVE E
z 9
Ua O
EXPECTED B E H AVIO R f
THERMAL DESIGN PENALTY
/
INCLUDED IN ORIGINAL DESIGN
/
0 100 6 E.E Co
s s.
l t
_ 26 I
i FIGURE 2.2 COMBUSTION ENGINEERING MODEL 0
DNBR e
ca ZQ CE CURVEg 9-i oD Cwc:
THERMAL DESIGN PENALTY INCLUDED IN INiTI AL DESIGN 0
100cb
_AC CO 1
I I
4 1
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