ML20107A935
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- n. c h 0-ios Lg Jersey Central Nonyf & Light Company (M.9 w
M A DitON AVENUE AT PUNCH 00WL RO AD e MORRi$ TOWN. N.J. 07950 e 539 6111 August 15, 197 h
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[1 M* Np Mr. Robert J. Schemel Directorate of Licensing U. S. Atomic Energy Cocrdssion
' Washington, DC 20545
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References:
(1) AEC Letter from R. J. Schemel, July 16,19 (2)
GE Topical Report, NEDM-10735, Supplement 6.
(3) JCPEL Co., Oyster. Creek Nuclear Generating Statian Facility Change Request Number 4 and Supplements.
Dear Mr. Schemel:
SUBJECT:
FUEL 1sENSITICATION In your letter of July 16,1973 (Ref erence 1), you requested tha t we provide the necesscry anajyr,es and other relevant data fer determininn the concequences of denstfication and the effects on normal operation, anticipated transienta and accidents at, the Oyster Creek Nucicar Generating Statica (OC)~ using the AFC nuidance z.t tached to that Icttcr. The letter ntated that if analyses indicate that changes in design or operating conditions are necessary to m.nintain rcquired mergins as a result of using the Staff guidance, we should submit. proposec' changes and operating limitations with the analyses.
Reference 2 has been prepared by General Electric as a generic response to the AEC's densification concerns, JCP6L Co. hat, reviewed this docu::'ent and finds it applicable to the fuel in OC.
It should be noted, however, that thn information cor,tained in Teble 4-1 associnted with Plant A (which is OC) ir not valid an it is representative of conditions not allowed by present Teehnical Specificat.f ons.
A revised version of this information is presented in Tabic 1.
This response includen the analysis of both General Electric fuel (Types I and II) and EyJr,0:? f uel ('Iypes III and IIIE) currently in the j
c ore. Analyses of the General Ilcetric fuel specific to Oc are contained in Mcference 2 (Section 6.3), whJle the analyses of the EXXON fuel are j
contained in Attachnant I of this letter. The results of ull analyses are sunuerized in Table 1.
Note that results have been prc,vided in the postulated Lone of Coolant AccJdent (LOCA) case for three cases including:
Ub'I
/
7 9604150200 960213 PDR FOIA j
DEKOK95-258 PDR
Mr.~ R. J. Schtmed
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Page 2 1
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- a.. The current Interim Acceptance Criteria (IAC) model b.
.The IAC model utilizing the AEC densification guidelines and applying the 95/90 confidence level to the gap conductance of the limiting rod c.
The IAC model utilizing the AEC densification guidelines and applying the 95/90 confidence level to the gap conductance for all rods.
The basis for these calculations and statistical confidence levels.
are discussed at length in Reference 2.
In. addition to the requested analyses, a proposed Technical Speciff-cation is presented in Attachment II.
This proposed Technical Specification allows for more direct monitoring of parameters determining the behavior
~ of fuel under postulated LOCA conditions.
I The assumed effects of fuci densification which have been considered in the analyses are the potential for (a) local power spikes resulting from axial fuel column gaps, (b) increased -linear heat generation rate
.due to pellet axial shrinkage, (c) cladding collapse at the location of axial fuel column gaps, and (d) reduced pellet-clad thermal conductance 4
due to increased pellet-to-clad gap as it may influence stored energy.
-The results of conservative analyses of local power spikes are presented.
The analyses yielded the conclusion that there is >95% confidence that no more than one rod in any existing fuel type will have a power spike
>5% in magnitude.
Further, due to the nature of the axial power distri-bution in a BWR, this maximum spike magni.tude will not occur at the limiting or highest power generating axial location in the rod.
Ihe results of the analysis of linear heat generation rate (LHGR) change due to densification show that the pellet axial shrinkage will be mor-e than offset by the effects of axial thermal expansion in the rods near the 11titing condition.
Thus, no effect on LHCR is expected.
The results of analysis of cladding creep collapse for existing LUR fuel types with more than one cycle of operation are also summarized. The 1
results show that creep collapse will not occur.
Evaluations were calcu-lated for existing fuel operating through September,1974.
Evaluations beyond this point will be conducted at a future date.
The additional major conclusions of the analyses are summarized below:
a.
Nornal Operation and Anticipated Transients:
The analysis of the nasumod densification phenomena and their effect on plant. safety _ bases have been examined.
It is i
concluded that the safety design basis criteria and the current i
l Techeical Specifient dee eith it.s im,no wd total power peaking
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Mr. - R.' J. Sch2mel Paga 3 I
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. f actor for both normal plant operation and' anticipated transients are still applicable. Additional power peaking due to power spikes calculated utilizing the AEC model are adequately accounted for in conservation in existing calculations. Other assumed der.sification phenomena have minor effects on normal plant operation and on the outcome of anticipated transients.
b.
Design Basis Accidents:
The' four design basis accidents (Control Rod Drop Accident,
-Main Steam Line Break Accident, Refueling Accident, Loss of Coolant Accident) have'been considered for effects due to the assumed densification phenomena.
It has been concluded-from these evaluations that the previously reported acceptable results for the Control Rod Drop Accident, Main Steam Line Break accident, and Refueling Accident are still valid.
Current analytical methods employad to evaluate Loss of 4
Coolant Accidents have been utilized to determine the effects i
of assumed densification phenonena. Analysis of small break
)
conditions indicate lower peak clad temperatures than previously reported._ Therefore, current small break analysis remain valid.
Tlie results of the design bases LOCA analyses (i.e., the recirculation line break) are significantly affected by the reduced gap conductance resulting from the utilization of the AEC model-in representing the postulated densification phenomena.
The effect is manifested in the calculated peak cladding temperature.
The value of peah cladding temperature is calcu-lated using both interpretations of the AEC densification model.
The utilization of either interpretation of the AEC densification model in conjunction with the proposed Technical Specification contained herein and additional constraints on the reactor power distribution requires the derating of the OC Nuclear Generating Station to meet the IAC.
The utilization of a 95/90 confidence limit on the limiting rod results in an approximate derate of 4% from current licensed power.
The utilization of a 95/90 confidence limit on all rods results in a derate of 11% from current licensed power.
Analyses-performed by both GE and EXXON, utilizing never, more sophisticated models than either interpretation of the AEC densification model indicate no required derating.
I k'ith regard to the postulated LOCA, it is again emphasized that the OC Nuclear Generating Station satisfies the IAC now applicable to this event with no additional restrictions on modes of operation.
It is rade ricar.in Reference 2 and Attachtent I that the principle effect of the Staff's fuel densification model is on the LOCA calculation and specifically a modification to the value of 1000 BTU /hr-ft2 F for pellet to clad gap n
conducta ce in the AEC approved noneral r.lectric ECCS evaluat Nn mdels utilierQ by JCFEL Co. fer OC fuci n
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- Mr. R. J4 Schemel
.Pags 4-
'New, sophisticated analytical models representing the changes of~ gap conductance in relation to the various phenomena. including fuel densifi-cation. associated with _ increasing exposure are 'in different stages of de velopmen t.
Such a model' developed by EXXON Nuclear and expected to be submitted to the AEC in September, shows that cdequate margin to limiting I
conditions now exist for both GE and EXXON fuel for the postulated design basis;LOCA.
General Electric is.now developing a similarly sophisticated model which -is expected to be submitted ' to the AEC by Decenber 1,1973.
The more icportant effects considered by these models and not now repre-4 sented in the AEC guidance are fuel pellet swelling, cracking and_ gap closure, and cladding creepdown.
Information on these phenomena has already been presented to the Staff.
It is the judgment of JCP&L Co. based on a review of all evaluation models.available and specifically the statistics which are the basis for the AEC guidance, that there is no safety reason for impicmentation of any additional operating restrictions, and that continued operation
- of OC under its existing license presents no undue hazard to the health and safety'of the public, Further, examination of the statistical analysis performed in
~
e Appendix A to keference 2 establishes as suitable the conservatism of j
the AEC densification guidelines applying the 95/90 confidence 1cvel to the gap conductance of the limiting ro'd.
However, it is the judgment of JCP6L Co. that even this interpretation of the availabic data is overly 1
restrictive in light of the aforementioned results of more sophisticated analytical models and existing experimental data aircady presented to the AEC.
Finally,_in response to the request for proposed operating limits consistent wita the results of the analyses, the following recommandations are'made:
n.
_It is recormended that a Techn4 cal Specification defined as a Limiting Condition for Operation and as described in j
Attachment II to this let ter bo implemented irnediately.
l-It should be emphasized that this recommendaticn provides
[
more direct nanitoring of LOCA limiting paraneters and is
}
consistent with present operating restrictions. Curve B of Figure 1 of Attachment II defines this lirdt.
b.
'In the event that the Commission finds it necessary to g
further restrict the operation of the OC reactor, it is recommended that this restriction be limited to that defined by the results of analyses utilizing the AEC densification guidelir.cs and applying the 95/90 confidence 4
0 7
g.
Mr." R. J. Schemel Page 5 1evel to the--limiting rod.
This restrictive mode of operation is shown for the CE fuel as Curve C'of Figure 1 of Attachment II.
Note that the Technical rpecification-proposed in a. above would be utilizea but with Curve-C as the limit line. Curves C' and C" representing the limit for the 95/90 confidence for the limiting rod Jassumption for the EXXON Type III 'and IIIE fuel will be
-supplied at a later date. Curves D, D' and D" represent the limiting conditions corresponding to the 95/90 confidence level on all rods for the CE Type II and.
EXXON Types III and IIIE fuel respectively.
c.
It is recommended that.the AEC continue its review of the
-fuel densification phenomenon and that the expected new analytical models for predicting the behavior of fuel under
. irradiation, including densification, and analyzing LOCA be reviewed and acted up,on expeditiously.
-We trust thct the information contained herein meets the requirements specified 'in your letter of July 16, 1973.
Please advise us promptly if.
any additional information is required.
i Very truly yours, p
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' I IvanR.Finfro'ck,p.
- v Vice President e
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Attachments t
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ATTACHIENT 1 EXXON l!UCLEAR OYSTER CREEK DErlSIFICAT10N Af!ALYSIS (CompliancewithAECJuly 16, 1973, Guidelines)
The analyses discussed below reflect calculations perfont.ed in response to the letter to Jersey Central Power and Light f rom R. J. Schemel dated July 16, 1973.
The guidelines given in Enclosure B to that letter for Exxon I;uclear fuel were followed in this analysis.
The results of the analysis indicate that if the specified directives are applied, the peak power density of the Oyster Creek reactor must be restricted to an axial x radial peaking factor product of 2.146 while operating at full power (1930 tiWth), or to a thernal power icvel reduced in inverse proportion to the axial x radial power peaking factor product ratioed to 2.33.
Only analyses of the type III E fuel which are affected by the assurced densification phenomena are addrcssed below.
Other analyses of the performance and safety aspects of this fuel are presented in Facility Change Request flo. 4, dated Janvery 18, 1973, and Supplements 1 and 3 thereto.
I.
DEllSIFICATIO'! EFFECT 0.'! I?OPMAL OPERATION The maxinun heating rate for the type III E fuel at rated power is as given in Supplement 1 to FCR lio. 4 as 17.2 kw-ft with the reactor at full power. Under these conditions, the liinimum Critical llent Flux Ratio was calculated to be 2.0 (XI!-1 CHF correlation).
The effects on these values of iraposing the densification criteria stated in the AEC letter of July 16, 1973, are as follows:
A.
Power Snike !% del Use of the power soil.e model pi esrented in Suppir=ent 3 to FCC !!o. 4 tesulted in a conclusion that, vith 95C ccnfidence, ne more than one fuel rod in the reacter v!ould eneed a power spite (due to axial peilet
l -
(
p r
columngaps)ofabout2%.
Utilizing this'same model, but applying the following equation to determine the maximum gap size:
0.965 al =
+ 0.004 L
2 0.965 - 0.9405,'.004 L
=
= (. 0162) L The maximum axial gap size at a pellet column elevation of 126 inches (above which the fuel power is normally so low that the added spike effect can be ignored) is calculated to be 2.05 inches.
Applying the same relative gap size distribution and gap frequency models indicated in Supplement 3 to FCR No. 4, it is calculated that, with 95% confidence, no more than one fuel rod in the reactor core would exceed a power spike s
.of abuul 4.6%.
B.
Linear Heat Generation Model Utilizing the AEC criteria in the July 16, 1973, letter, the decrease in fuel colunn length (and conversely, the result of this effect on
.the LHGR) is calculated by:
0.965 - ni' L
el = -
2 L
=
.012 L. or t.LilGR = +. 012 LHGR Compensating for this offect, but not acccunted'for in the original analysis for the type Ill E fuel, is the thennal expansion of the fuel i
- l.0:liTQl lK'llet, dille' i th'lb;t"bly average)
(
)
- at power, which increases the fuel pellet axial length, compared to the cold manufactured length.
At a linear heat generation m.te of 17.2 kw/f t, the axial thermal expansion of the fuel pellet column (utilizing the calculated U0 temperature at the inner edge of the pellet dish).is +1.2%, which is 2
reflected as a reduction in LHGR below the design evaluation value of -1.2%.
The engineering heat flux subfactor due to pellet density variation on an average assembly basis is:
p +o
.9405 +.0055 1.006
=
=
P
.9405 The calculation of the average and maximum LHGR in the reactor core depends upon (ameng other facters) the total active length of fuel pellet columns among which the thermal powcr load is shared.
The specified length of the pellet columns in Exxon !!uclear fuel (dold) 1s 144 1 0.25 inch, or i.17%.
All fuel rod rellet columns were verifed to fall within this manufacturing tolerance.
It is assumed that the average length of all the fuel pellet columns is 144 inches, and no adjustment in the average or maximum LiiGR is made to compensate for deviations from this nean.
C.
Stored Energy Model Applying the directive given in the AEC letter of July 16, 1973, the radial gap coefficient is calculated to be 500 DTU/hr F z t the 95/90 confidence level based on the empirical gcp coefficient correlation with the following attributes:
L*
3
(.-s.
4 I
e LHGR.='17.2 kw/ft; e
pi-
= 93.5% of theoretical density
.234 inch.(manufactured) e 2a
=
i Applying this value!and the UO thermal conductivity data of Lyons et al, 2
-the calculated maximum pellet centerline temperature at 17.2 kw/ft is 4620 F.
This temperature is well below the 5030 F melting point of UO fuel.
The. effects of operational transients on the fuel centerline-2 temperature is discussed 'in Section 11 below.
The effects of fuel exposure and gadolinia burnable poison addition are as described in L
FCR I o. 4. January 18, 1973.
D.-
liet Densification Effect on llormal Operation The. net effect of densification on normal operation of applying 'the assumed densification effects discussed above is to increase the cssumed local peaking factor used to estab'lish the maximum LHGR for core moni-toring purposes.
This increase is:
Present Analysis Previous Analysis Power Spike 1.046 1.000 Decrease in column length due to densification 1.012 1.000 Fuel column thernal expansion
.988 1.000 Engineering heat flux factor 1.006 1.016 Total 1.052 1.016-Increase in hot spot factor in present analysis-1.030 9
L m
e
5 h
)
r.
Thus', a peaking factor '(F ) of 1.036 should be superimposed on the core D
- monitoring factors for the purpose of determining compliance with the Technical Specification limit of 17.2 kw/f t.
As indicated in (C) above, this limit provides' sufficient margin to insure operation below the U0 mciting p int and MCllFR limit during normal operation and anticipated 2
Maintaining the maximum L11GR at 17.2 kW/ft also maintains' the calculated MCHFR at 2.0 for the same assumed core thermal hydraulic conditions given in FCR lio. 4, January 18, 1973.
II. ' dells!FICAT10tl EFFECT O'l TRAi!SIENT C0liSEQUEllCES
.. Sensitivity studies assuming undensified fuel indicate that,.for the worst case of a transient involving a significant power spike (a turbine trip with-out bypass), the peak fuel pellet temperature increases about 100 *F.
Considering the effects of assumed densification on the maximum steady-state U0 temperature indicated in Section I, C, above, a.large mergin.to 2
the U0 melting point remains during this. transient.
2 Ir,fpection of the results of the rod withdrawal incident analysis presented
-in FCR fio. 4 shows that the APRi4 rod block setting will. limit the local-power density increase to less than 15%.
This increase in the maximum steady-state peak power would' increase the peak pellet temperature from about 4620 *F to 5200 *F, which exceeds the U0 melting point.
To control 2
this peak ten,perature to less than 5080 *F, the maximum local rod heat flux would have to be reduced about 3%.
Since the results of the LOCA analysis (Section 111) require a power reduction in excess of this value, the LOCA analysis results are controlling.
1
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111. :DENSIFICATION EFFECTS ON' ACCIDENT CONSE0VENCES The calculated consequences of the Main Steam Line Break accident, the Refueling accident, and the Rod Drop. accident are not changed by the effects of-the assumed densification guidelines.
The effects of the AEC directives of the July 16 1973, letter on the calculated Loss of Coolant accident consequences are as follows:
A.
Poser Spike Model The effect of pellet column gaps on the results of the LOCA analysis has been considered.
In the limiting case, two competing phenomena occur:
A pellet column gap in a given fuel rod results in an increase in a.
the heat generation rate of adjacent rods.
For BWR's', the worst case is a gap in a fuel rod adjacent to the rod which reaches the
~
maximum clad temperature during the LOCA.
b.-
The presence of a pellet column gap in the axial plane of interest results in a reduction in bundle power by about 2% ('the power fraction of the " missing" rod segment) for that axial plane.
Sensitivity studies of a limiting case, as described in (a) and (b) above, indicate that the net effect of pellet column gapping on the LOCA analysis is to slightly decrease the' calculated peak clad temperature.
-Hence, this phenomena is disregarded in this LOCA analysis.
B.
Linear Heat Generation Rate Model Utilizing' the AEC directives in the July 16, 1973, letter,-the decrease a-w
1
.(
)
c,
, in fuel length (core average *) for the purpose of.LOCA calculations-isJobtained from:
0.965 - pi' AL = -
L 2
The mean pi?for this type III E fuel was reported in Supplement 3 to FCR No. 4 as 94.05% TD.- Hence, AL/L = -.01225.
The core averaged fuel, pellet thermal expension is calculated to be about. 0.5% at a core-averaged fuel power of 5.2 kw/ft.
Hence, the net effect of densification on the LHGR for the purpose of LOCA calculations is +.012
.005 = +.007 or +.07%.
C.
Stored Energy Model The AEC directive of July 16,1973, to utilize reported gap coefficient
. values to obtain a gap coefficient model as a function of LHGR, gap size,:and pellet diameter that predicts the data with 955 confidence j
that 90'/: of future events will exceed predictions was followed. Table (1) is.a summary of' all applicable gap coefficient data at beginning of life with helium fill gas used to construct the gap coefficient model.
The observed gap. heat transfer coefficients, ocliet diameters, and gap-to-diameter ratio used in the model development are prpsented in Table (2).
The form of.
the gap coefficient model assumed to describe the data was based on the form of the analytical solution for heat transfer with cylindrical geometry.
The, empirical model has the form:
o h = a (LHGR)al (9/D)a2 (D)a3 o
g Assui.iing the core is loaded with type III E fuel.
- 8 ~-
{,.
)
r.
where 2
hg _= gap heat transfer coefficient, Btu /hr.-f t,.p.
LHGR = linear heat generation rate, kw/ft I
g/D = cold =diamettial gap-to pelle, diameter ratio
- The parameters of the empirical model representing the mean of the data were evaluated by a least square-analysis from Table II.
Since the AEC directive specified that a 95/90 lower tolerance
' limit accompany the gap coefficient model, a statistical evaluation of the difference between the predicted and the reported gap coefficient c
was performed.
It was found that multiplication of the empirical gap coefficient for_ the mean of the data by 0.826 yields a predicted gap coefficient value that satisfies the 95/90 lower tolerance limit criteria.
~ The resultant gap ~ coefficient model for. 95% confidence that 90% of-future events will exceed prediction is h = 14.98 (LilGR)o.374 (g/D) o.523 (D)~O 94 J
g Figure 1 provides a comparison of the prediction with the 95/90
~
empirical gap coefficient model and the reported gap coefficients All i-of the reported gap coefficients except a single point. from Reference (4) l at 7.5 kw/f t, a g/D = 0.0326 and reported gap coefficient of 473 Btu /hr.-f t
,,7 2
l' are shown to be underpredicted in Figure 1.
Instantaneous densification of the U0 fuel pellets in the type 111 E 2
{
fuel is assumed.
Since the LOCA calculation results are sensitive to i
l the bundle-averaged stored energy, the pellet diameters and gap sizes
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are calculated assuming the lowest mean value of the individual pellet-lot density.is used to determine the densified pellet diameter. As-reported in Supplement 3 to_ FCR No. 4, the lowest mean density va'lue of an individual pellet lut was 92.3% TD.
Hence.
0.965 - 0.923' A r/ r = -
= - 0.014 3
j The cold fuel pellet OD is adjusted using this value and a " densified" pellet-to-clad. cold gap obtained.
The curves derived from fits to the experimental data described above were used in conjunction with the
'" densified" cold gaps -to obtain gap conductivities for each fuel rod.
These values are then used to calculate-the peak clad temperature during the LOCA as described in FCR No. 3, January 13,,1973.
The results of this calculation are presented in Figure 2, where the peak i
clad temperature is plotted as a function of the product of the axial i
x radial peaking factors, with the reactor power assumed to be 1930 Muth.
i This figure indicates that the peck clad temperature is 2300 *F when the axial x radial product is 2.146.
The radge of gap coefficients for several values of this product is as shown on the figure.
The above calculation conservatively assumes maximum densification j
to occur instantly although it is expected that full densification would not be accomplished until several days or. weeks have elapsed.
During this period gap closure as a result of pellet cracking would tend to compensate for a reduction in gap. coefficient due to densification.
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)
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REFEREtiCES 1.
- h. S. Bain, Plicroccopic, Autoradicgraphic and nect/ sheath Heat Transfer Studies en UO htcl Elmento, hECC-2588, Junc 1966.
a g
2.
R. fl. Duncan, Rabbit Capsule Irradiation of UO, Tcuminal Report, g
CVilA-142,1962 3.
G. K$aerhein and E. Roistad, In-Pile Determination of UC,, Thcrnal Conductivity, Density Effects and Gap Conductancc, HPR-80.
4.
D. C. Ditmore and R. B. Eikins, Dancification Consideraticno in BWR Fuel Design and Performance, tied'1-10735, December, 1972.
4 i
I i
i j
Table 1 SUMi%RY OF PHYSICAL PARAMETERS OF DATA USED DEVELOPMEiiT OF EMPIRICAL GAP COEFFICIEf1T MODEL AECL-2588(1)
CVi!A-142(2)
HPR-80(3) f4) flEDM-10735 O
Fellet Dianeter, in s.65 s.43
.49
.488
~
Diametral gap, in
.02 i
.0256
.0066
.016 Linear Heat Generation Rate, kw/f t 25.3 18,24 2.8 - 15.0 17.5 - 20 Exposure, MWD /NTM sO sO s0 s0 Fill Gas He He He He,
Reference Temperature Melting Equiaxed Thernoccuple Equiaxed Cladding Material SST Zr Zr Zr
% Theoretical Censity 97.3 94
-96 95.7 External Pressure, psi 100 406 s 1000 Cladding Thickness, in s.026
.0218 0.030
1 Table 2
+.
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I-SUltlARY OF DATA USED FOR DEVELOPMEliT OF EMPIRICAL GAP COEFFICIENT 1100EL Gap Coefficient;-
LHGR Pellet 0.D.,
2
_ Btu /hr-f t..F.
kw/ft
. g D, '
inch Data Source
.705 17.5
.0326
.488 GE-NEMO 10735 550
'17.5
.0326
.488 GE-NEDM 10735 555 17.5
.0326
,'. 4 88 GE-NEDM 10735 473 17.5
.0326
.488 GE-NEDM 10735 10101
'20.0
.0326
.488-GE-NE0*4 10735 623 20.0
.0326
.488 GE-NEDM 10735 I
486 25.3-
.0404
.650 AECL-2588 486 25.4
.0404-
.650
.AECL-2588
~
570.
18.0
.0590
.430 CVNA-142
'520 24.0
.0590
.430 CVNA-142 565-2.83
.0133
.490 HPR-80
'580
'3.66
.0133
.490 HPR-80 600-4.57
.0133
.490 llPR-80
-625 5.7f
.0133
.490 HPR.
700 7.99
.0133
.490 HPR-80 720 8.72
.0133
.490 HPR-80 74 0 8.,99
.0133
.490 HPR-80 750 9.17
.0133
.490 HPR-80
-760 9.57
.0133
.490 HPR-80 830 11.37
.0133
.490 HPR-80 1050 14.97
.0133
.490 llPR-80
~
600 4.48 0.135
.490 llPR-80 620 5.36 0.135
,.490 llPR-80 630 5.70
.0135
!.490 llPR-80 670 7.16
.0135
.490 HPR-80 720 8.56
.0135
.490 liPR-80 730 8.841
.0135
.490 HPR-80 750 9.39
.0135
.490 HPR-80 1
755 9.54
.0135
.490
- IIPR-80 800 10.55.
.0135
.490 flPR-80 i
.810 10.79
.0135
.490 llPR-80 j
860 12.04
.0135
.490 llPR-80
- 880~
12.47
.0135.
.190 IIPR-80 l
l
~-,
n
+
Tabic 2(Continued) 910 13.11 0135
.490 HPR-80 940 13.50 0135
.490 HPR-80 970 14.11 0135
.490 HPR-80 O
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FIGIRE 2 0YSTER CP.EEK TYPE IIIE FUEL Per AEC Letter.to GPU, July 16,:1973
~
2800 -
'?
s J
s
{ 2500 -
430 - 512
. 5 GAP COEFFICIENT D
RANGE h
O 2400 424 - 496 d
~~-
-~~~~~~~
~
E 2300-1 t'
414 - 485 g
13 2200 1
3 403 - 479 I
L' l
390 - 469 1
2000
-I 1
I I
i 1
i i
1.6 1.7 l3 1.9 2.0 2.1 2.2 2.3 2.4 1800 Product of Axial and Radial Peaking Factors 01930 fir!t i
9'0% t 10'0 /,
7'0; 30, 92.1%
Percent of Tripic Product 2.331930 IMt 1
2 e
m
___t___v_a_-,
w
____.-,,.r-
(
I APPENDIX A
~
RIEULTS OF EXXON ANALYSIS FOR TYPE III FUEL i
CURRENT ANALYSIS:
g PCT AFLHGR RxA Power 1000 2225 13 3 2 33 1930 Using AEC Guidance of July 16, 1973:
410-475 2601 13 3 2 33 1930 400-456 2300 12.2 2.20 1850 Power Spike Analysis See Discussion for IIIE Transient Analysis See Discussion for IIIE e
e i
l
)
i 1
1
G
)
ATTACRMENT II PROPOSED TECHNICAL SPECIFICATION CHANGE 1A Specificatior. to be Changed Section 3. Limiting conditions for Operation IB Extent of Change Add Specification 3.10 to Section 3.
Add Figure 3.10.1 1C Change Requested 3.10 AVERAGE PLA'iAR HEAT GENERATION RATE
,Applicabili t;g: Applies to the monitoring of the mea:imum average planar linear heat generation rate (MAPLHCR)
Objective:
To limit the APLHGR in such a manner as to conform to the peak clad temperature limitations during a postu-l'ated loss-of-coolant accident as cpecified in the Interim Acceptance Criteria.
Specification: The average linear heat gene: ration rate at any axial cross section of any fuel bundle in the core (Average Planar Linear Heat Generation Rate, APLHGR) shall not exceed the operating icvel (MAPLHGR) shown by Curve B at. Figure 3.10.1 (See Attached Figure!3.10.1)
Bases:
To be provided subsequent to AEC Staff evaluation.
(Note: The justification for choosing Curve B of figure 3.10.1 is given in the cover letter to this submittal).
C i
k 2A
_ Specification to be Changed Section 4.
Surveillance Requirements
+
2B Extent of Change Add Specification 4.10 to Section 4.
2C Change Requested 4.10 AVERAGE PIJNAR HEAT GENERATION RATE Applicability: Applies to the surveillance of the average Planar Linear Heat Generation Rate (APLHGR).
Objective:
To assure thct the APLHGR is within,the limitations imposed by Curve B of Figure 3.10.1 Specification:
Daily during reactor operation, the maximum Average Planar Linear Heat Generation Rate shall be estimated and checked against Curve B of Finure 3.10.1 and adjusted if required.
\\
\\
Basis:
The peak clad temperature which may result in the event of a postulated loss-of-coolant accident j
varies proportionally to the MAPLIIGR. Daily surveil-lance of the MAPLilGR will assure that the LOCA peak clad temperature will conform to the limitations imposed by the Interim Acceptance Criteria.
.