ML18088B190
| ML18088B190 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/30/1978 |
| From: | Combustion Engineering |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18088B190 (137) | |
Text
'.', 'EN-89(F) -NP Increased Water Hole Peaking in Operating Reactors (St. Lucie-1)
March 30, 1978 Combustion Engineering, Inc.
Nuclear Power Systems Windsor, Connecticut
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LEGAL NOTlCE
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THIS REPORT IVAS PREPARED AS AN ACCOUNT OF IiVORK SPONSORED BY COf BUSTION ENGINEERING, INC.
NEITHER COMBUSTION ENGINEERING NOR ANYPFRSOiJ ACTliJG ON ITS BEHALF:
A.
MAKES ANY (VARRANTY OR REPRESENTATION, EXPRESS OR NPLIED INCLUDING THE V(ARRANTIES OF FITNESS FOR A PARTICULAR P
RPOSE OR'iERCHANTAOILITY, 'VJITH RESPECT TO THE
- ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORViATIONCOiVTAliJED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION;-"APPAIEATUS,ViFTHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OL+JED RIGHTS; OR B. ASSUMES ANY LIASILITIESVJITH RESPECT TO THE USE OF, OR FOR DAMAGES RFSULTING FROM THF US OF, ANY INFORMAL'IATION,APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
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CRITERIA FOR PROPRIETARY'HFORHATIOI<
Code numbers 1 to 6 have been used in the margin of the text to identify proprietary information..The-fol-lowing list explains the"criteria associated with these code numbers.
l.
Information reveals confidential cost or price information, commercial strategies, production capabiltiies, or budget levels of Combustion Engineering, Inc., its customers or suppliers.
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The information reveals data or material concerning Combustion Engineering or custo...er funded research or development plans or programs of substantial present or potential comp'etitive advantage
-to Combustion Engineering, Inc.
3.
The use o
the informat',on by a competitor would substantially decrease-A...expenditure,-in tim or resources, in designing, producing or marketing a similar product.
4.
The information consists of test data or other similar data con-cerning a process, method or compon nt, the application of which results in a substantial ccmpetitive advantage to Combustion Engineering, Inc.
5 T le
~el Gl i ia I.; on reveals special aspects of a process,
- method, component or the like, the exclusive use of which resu]is in
~ a substantial ccmpetitive advantage to Combustion Engineering, Inc.
6.
The information contains ideas for which patent protection is likely to be, sought.
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Intro due tion This document contains the information presented to the= NRC on December 16, 19??
and January 18, 19?8.
Section 1 presents the results of investigations of pin power peaking in assemblies with CEA waterholes using a newly developed multigroup transport theory calculation method.
These results indicate that the current standard design model is underpredicting the power in fuel pins adjacent to CEA waterholes by approximately 4.5/.
Section 2 describes conservatisms in the analysis package used to accommodate
~the increased water hole peaking for first cycle of St. Lucie Unit l.
Explicit use of the methods described will be employed in reload designs starting with Cycle 2.
Sections 3 and 4 further describe improvements in thermal margin/low pressure trip synthesis and a partial credit for TORC/CE-1 for DNB LCO respectively that are employed in the ana'lysis package discussed in Section 3.
V
4 Table of Contents Section 1
Section 2
Section 3
Section 4
Hew Peaking Itodel Actions to Accommodate Increased llaterhole Peaking for St. Lucie Unit 1
Improved Thermal Hargin/Loiv Pressure Trip Synthesis Partial Credit for TORC/CE-1 in ASI t1onitoring Limits
~r 0
New Peaking Model Slide 1
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The purpose of this presentation 's to present the results of recent investiga-tions of pin power peaking in assemblies with CEA waterholes using a newly developed mul tigroup transport theory calculation for the uel assembly.
These
- results, which include a
nevi evaluation of the available experimental data
- base, indicate that the standard design model is underpredicting the
. power in pir adjacent to.CEA waterholes by about 4.5>.
ate'will'lso describe how the increased pin peaking will be incorporated in the desi gn power di s tribution ca 1 cul ations..
He will also present here an evaluation of the effects of.the increased peaking on current operating plants.
Me do not plan to change the methods used on these plants now, but rather will incorporate the new methods in future reload submittals.
h'e will show sufficient margin presently exists to cover these effects and that there are no safety concerns with continued operation of these plants.
Slide 2.
The new peaking model will be incorporated in the overall setpoint analysis along with the use of a reduced. power distribution measurement uncertainty using the values justified in C-E Topical Reports on this subject (CEt<PO-153 and CEHPD-145).
An evaluation of the current cycles of all C-E operating plants (except Haine Yankee and Palisades)'hows
'that the increased pin peaking can be account d for
'ith available marg.ins in the overall setpoint analysis.
There are no sa ety considerations or power capability restrictions and no significant technical specification changes are required.
As we will describe later, a small (0.5%)
additional conservatism will be added to the monitoring limits for peak YM/FT.
An evaluation of the increased peaking will be shown for each cycle as the action requirement depends on the power distribution uncertainty presently in effect on the current or upcoming cycle.
Slide 3
For purposes of the safety analysis and povier distribution measurement uncertainty assessment, the total 3-D nuclear peaking factor is constructed from 3 components.
The fuel assembly axial peaking factor (Fz) is obtained from the normalized axial shape in each.fuel assembly and its uncertainty is evaluated from signal
, reproducibility and the precision.of the axial fitting technique used to construct a continuous distribution from the 4 rhodium incore detector segments.
The radial power distribution -(FR), which represents the average fuel ass'enibly
-axially integrated po>>'er, has an uncertainty component that is evaluated from'omparisons of calculated and measured incore instr ument signal radial distri-butions.
This is the uncertainty component that makes use of reactor operating data.
The third factor is the maximum pin power to fuel assembly average power
'nd is the subject of this presentation.
Since the incore instrument measures instrument reaction rates rather than the desir'ed maximum pin power, it is necessary to rely on calculated factors to infer the pin power in an operating reactor.
The uncertainty component associated with these calculated factors
's evaluated from comparison of calculated and measured fuel pin relative powers in separate critical experiments.
We have used such critical experiments to determine the uncertainty associated with use of the new pin power peaking m.odel.
Slide 4
The incore instrument system in C-E reactors was initially installed
~o provide general information on the details of the cr ro. rn'"er ~istrihution and was not used to demonstrate.
compliance with any Technical Specification operating limits.
An un-certainty of 10" was ol"lginal'ly assigned by HRC for the initial use of the system.
With the introduction of very low KW/FT limits associated with the new LOCA requirements, it became. desirable
= to make direct use of the incore instruments to demonstrate plant operation within these limits.
C-E submitted a repor t to HRC in late 1973 justifyinq an uncert'ainty of 6.5X when using the incore instruments for this purpose.
NRC assigned an uncertainty of 8'which was
.the highest for all PWR vendors.
Since this time, C-E has submitted detailed topical reports on both the basic uncertainty of the fixed instrument system (CENPD-153)'nd the uncertainty associated
>>i th use of the C-E incore instrument analysis code CEtlPO-145.
These reports are still being reviewed by tlRC.
Slide 5
.A continuing effort has been underway at C-E to improve all aspects of the physics design models including pin power peaking, rod worths, temperature coeffi-
- cients, and reactivity (boron) rundown.
In 1968, C-E performed a set of critical experiments with measurements of pin power peaking and rod worths and assemblies with CEA waterholes.
The evaluation of pin peaking and rod worth showed good agreement with design models in use at that time (approximately
[
5on pin peaks near CEA waterholes).
Subsequent effort then focused on im-proved agreement on reactivity rundown and temperature coefficients.
The detailed treatment of the
- -. CEA waterholes has a large impact on these quantities as well as. on the pcwer peaking in adjacent fuel pins and changes made
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3 0
here can impact ttre predictions of pin pea) inq I
In 1974, additional cr.itical ex)>crime>>ts were performed for UOp and mixed.
oxide latt.ices.
These experiments included t)ie first hot conditions and design model comparisons slrifted to this newer set of experiments.
These analyses,.
using the most, recent irrproverr nts in tlie desigrn model, did suqaest a
sr.",all
'(approximately[
Q underprediction of poi:er in pins near GEA i aterhoies; T}ese compo> isons are reported in CEl.'PD-153; this small un"erprediction was included as part of the overall uncertainty.
One of the
)))AC questions on CE)sPD-153 questioned irhether or not the underprediction at waterholes should be treated separately as a bias rather than as pal t of tire overall uncertainty, since the peak power. pins were expected to be adjacent to CEA waterlroles.
A new version of the OIT code, which contains an 85-group transport theory calcu-lations of the entire fuel asser~bly i>>eluding tiie detailed interaction between waterholes, power pins, and poison shirrs, lras now been applied to the prediction of waterhole pea).ing.
These first results iridicated t)iat the current design'odel underpredicted pin peakinq near CCA waterholes by amounts considerably larger than indicated by the comparisons wit)> the 1970 experiments.
In order to resolve this, the 1968 experiments liave been reanalyzed with tlie standard design model and with DIT.
An evaluation with the full data
- base, sliows that OIT over-predicts pin peal'.ing by [
]and the standard design niode1 underpredicts pin peaking by r) to Sr,.
These results led to the current action (described on the following slides) to account for this effect.
I Slide 6.
This'lide outlines the procedure that, will be employed to account for higher pin peaks adjacent to Cfh waterholes in ihe standard design models.
The standard design diffusion theory model (POg) is used for quarter core pin-bJ'-pin calcula-tions of the power 'distribution, POg can also be applied to calculations of specific fuel assemblies using appropriate operating conditions.
The transport theory DIT calculation can'nly be run for the assembly geometry.
Hhile it is possible to adjust, diffusion t)reory constants so that pin povrer peakinq will match trans-port theory results, we have not yet developed a prescription that will provide
. the simultaneous prediction of power peaking, temperature coefficients and reactivity rundown in diffusion theory to the desired accuracies.
For this reason, our oresent plan then is to use the transport theory DIT code to define a pin peaking bias that
'ill be applied to the results of the standard design model calculations..
This will be accoup7i shed in two steps; first, DIT is used for an analysis of the critical lattice experiroents to define a calculational uncertainty and bias for t)re best method.
In the second
- step, the different fuel assemblies (varyinq enrichment, poison pin loadinqs, etc) are analyzed with both DIT and t)>e standard desirln PDg models arid t)re difference determined for each type of fuel asserrbly.
These results are then combined with ttre results of the first step to produce a bias and uricertainty for the design method.
I4 0
,0 A consistency check of this procedure has been made by direct evaluation of the critical,,experiments uii.th the standard design model.
As expected, the uncertainty and bias established from this approach is comparable to that obtained from the described procedure.
=The bias is applied to the desiqn model calculations for subsequent use in the safety analysis and qeneration of coefficient libraries for tbe incore instrument power distribution monitor',ng conditions.
Slide 7
The results of application of the above. procedure for the high enrichment
'.n a typical C-E reload batch is shown here.
These results are obtained by depleting an assembly calculation usinq soluble boron representative of the core variation with cycle depletion.
As can be seen on th
- slide, DIT gives a pin peak of aboutL.
/above the standard POg model.
As stated earlier, DIT is expected to overpredict the pin peak by abouts
. This bias is sub-tracted from the upper curve to obtain the (approximately 4.5X} bias to be applied to the design PDg.
In this case, the difft.'rencebetween POg and DIT remains fairly constant throughout the cycle as a result of compensation between burndown of the pin peaks and reduction of soluble boron in the waterholes.
Slide 8
I Results of similar calculations are shown for C-fuel.
In this.
- case, the similar bias has a slight downward trend with burnup in the first cycle and.when the soluble boron level is increased to a value typical of beginning of second cycle, the bias shows a step decrease of about 1'f,.
This is a result of a smaller difference between the standard POg model and the DIT transport theory calculation when the waterhole contains a large amount of pol soll.
Similar curves will be generated for each specific fuel assembly in C-E reactors.
Slide 9
This slide is another presentation of the procedure described on Slide 6.
slide 10
'0 The DIT code includes an 85-group*neutron spectrum calculation with a spatial geometry that accounts for interaction effects between waterholes, fuel
- shims, and poison pins.
The assembly spatial calculation is based on integral transport theory and includes an explicit representation of the fuel pin and its surrounding moderator.
The fuel pins are not homogenized and a detailed mesh structure is used inside the pins to fully account for the details of the flux distribution within the fuck pin and its surroundinq moderator.
A
e Several approximations are employed in DIT to minimiz o minimize computer storage requirements and to achieve reasonable runninq times Th mes.
ese approximations were evaluated'nd justified through comparisons with exp]icit Honte Ca'rlo cal-culations.
- However, they will introduce small inacc b'
reasonable to exp ct DIT will overpredict pin peaking near waterholes.
This is what is found in the evaluation of critical experiments.
Slide ll This slide provides the characteristics of each of th t'
and the number of pins adjacent to waterholes that wer m
c o
e cr>> ca experiments experiment.
It is noted that a
ia were measured in each is no e
that a range of enrichments, volume fractions tempera-ure an boron PPl) 1ev 1
b l
'e s, have been covered.
)lhen account is taken of the 9
difference in moderator temperature, C-E,experiments
'.53 d "'5 a,56 a
.y n
o uranium ratio found in the C-E operating lattices.
The CNWL lattices have voluae ratios c~nsif)erahlv oreater than enc t
d i th C
a ices and so have been excluded from the data base fo'r evaluation of DIT biases and uncertainties.
Slide 12 s
Comparisons between the HORSE Honte Carlo code, the DIT integral transport
- code, and the'DOT discrete ordinance transport
- code, are shown for a typical. C-E fuel assembly with poison shims.
It is noted that, on the average, DIT gives a small
.;Ioverestisnate of the flux in pins adjacent to CEA waterhoies.
Slide 13 A comparison of measured and calculated pin peakinq for the Kritz lattice experiment is shown here.
The measurement and calculations are normalized to one over the central 14xl4 fuel pin array in the experiment; DIT shows an average overprediction of[
]for pins adjacent to CEA waterholes.
Slide 14
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This new information will be used to update the pin pea) ing uncertainty evalua-tions in Chapter 4 of CEHPD-153 to arrive a
a new overall uncertainty for use of the incore instrument system in measuring core power distributions.
These new results are based on an enlarged data
- base, including the C-E criticals from 1968 as well as the 1974 Kritz experiments.
The basis for the evaluation of pin peaking uncertainty will also be modified to include only those pins adjacent to CEA waterholes.
This directly responds to an WRC question on CEt)PD-153 and provides an improved basis because the peak pin in the core is expected to occur adjacent to a
CEA waterhole.
Slide 15 The DIT evaluations of the critical experiments are shown here.
The evaluation of all of the experimental points provides a bias (overprediction) of) and a 95-95 confidence level uncertainty of[
g;the uncertainty is less than that previously reported in CENPD-153.
The variation of the bias for the
v lume six sets of critical experiments can be correlated with chanqes in f 1 t t
~ nqes in ue o water vo ume ratios.
The larger biases shown for C-E f53, C-E F56 and Kritz are closest to the volume ratios encountered in the C-E reactor lattice.
- However, an average of all the experiments has been conservatively:chosen for applica-tion to the standard design model.
RRL: kf
PURPOSE IS TO DESCRIBE I'
~ RESULTS OF RECENT INVESTIGATIONS OF POWER PEAKING IH ASSEMBLIES WITH CEA MATERHOLES.
'IMPLEMENTATION OF HEW i~lODEL.
EYALVATION OF OPERATING PLANTS;
IMPLEMENTATION OF HE'rl t~ODEL INCORPORATE NEM PEAKING MODEL IN OYERALL SETPOIHT ANALYSIS COHCURREilT >lITH REDUCFD POWER DISTRIBUTIOt(
HEASVREHEHT UNCERTAINTY (5.2X on F; 4.6X on F).
EVALUATION OF OPERATING PLANTS (CURREiNT CYCLES)
~ ACCOUNT FOR EFFECT ON OYERALL SETPOIHT AiNALYSIS WITH AYAILABLEMARGINS.
HO SAFETY CONCERNS OR POh'ER CAPABILITY RESTRICTIOHS-HO SIGNIFICANT. TECHNICAL SPECIFICATION CHANGES RE(UIRED.
~ ACTION DEPENDS ON PO'tlER DISTRIBUTION UNCERTAINTIES IH EFFECT ON CURREHT OR UPGO>PING
- CYCLE,
FIXED INCORE SYSTEM UHCERTAI i'lTY F
CONSISTS OF THREE COi'1PONEHTS F
= FZ
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FR FP Fz ", FUEL ASSEtlBI Y AXIAL PEAKING FACTOR UNCERTAINTY COMPONENT EVALUATED FROtl SIGNAL REPRODUCIBILITY AHD PRECISIOiN OF AXIAL FITTIi'lG TECHNIQUE F
AVERAGE FUEL ASSEMBLY INTEGRATED POWER - RADIAL R
'UNCERTAINTY COtlPOHENT EVALUATED FROM COtiPARISOH OF CALCULATED AHD t'EASURED IHSTRUtlEHT SIGNAL RADIAL DiSTRIBUTIONS (AXIALLY INTEGRATED)
FP -
flAXItlUMPIN PO';IER TO FUEL ASSEMBLY AVERAGE"POWER
~ UHCERTAIHTY COMPONENT EVALUATED FROM COMPARISON OF CALCULATED AHD t~iEASURED FUEL PIH RELATIVE PO'<<ERS IH SEPARATE CRITICAL EXPERItlENTS
ftISTORY OF PO';,'ER DISTRIBUTION UNCERTAINTY 1972 LATE 1973 AUG. 1974 OCT. 1974
~ HOY. 1974 i
APRIL 1975 EARLIEST PLANTS (PALISADES AHD MAINE YANKEE) STARTED WITH lOX OH TOTAL PEAKIiHG FACTOR.
INTERIM REPORT TO HRC JUSTIFYING 6'.5~ -
HRC ASSIGNED 8Ã - HIGHEST FOR ALL PttR VENDORS.
C-E STATED INTENT TO SUBMIT TOPICAL REPORT (CEHPD-153P)
OH BASIC UNCERTAINTY OF FIXED IHSTRUttEHT SYSTE11 FOR MEASURIHG POilER PEAKING - JUSTIFIED 5.2~ 95-95 UNCERTAINTY ON F
q HRC RESPONDED BY RE(UESTIHG TOPICAL REPORT ON IHCORE IHSTRUtKHT ANALYSIS CODES OF ALL PNR YENDORS.
C-E SUBtiITTED CEt<PD-153P C-E SUBMITTED TOPICAL REPORT (CEHPD-145)
Ott INCA - USING INCA F I
UNCERTAINTY OF 5.7~ JUSTIFIED.
Ee.
1975 CEiHPD-153
<RCH 1976 CEHPD-145 HRC MADE EXTENSIVE RE(UEST. FOR ADDITIOiHAL INFO RYE IM UP TO THIS TIiME, OTHER P'rlR YEHDORS EITHER HAYE HOT SUBMITTED REPORTS OR HRC DID HOT PLAN TO REVIElf.
JULY 1977 C-E SUBt)ITTED RESPONSES TO NRC QUESTIONS QH CEHPD-145.
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C-E'SUBMITTED UPDATE TO CENP0-153, SECTION 3.2.4-DETECTOR DEPLETION.
DESIGi'f MODEL EVOLUTION 1968
-t1EASUREtfE,'<TS OF PIi't POHER PEAKING AND ROD WORTH PERFORt1ED FOR ASSEHBLIES lv'ITH CEA MATERHOLES.
-PEAKING AND ROD 'r'ORTH AGREEHEHT 1'lITH DESIG(l MODELS 'klAS GOOD.
-SUBSEQUENT EFFORT FOCUSED 0 f It1PROVIHG AGREEHEHT OH REACTIVITY RUt4004H AHD TEHPERATURE COEFFICIENTS. (MATERHOLE TREATt1EHT HAS LARGE If1PACT).
1974
-ADDITIONAL CRITICAL/EXPERIi1EHTS PERFORtfED FOR UO~ AHD MIXED OXIDE LATTICES.
-ANALYSIS SUGGESTED St1ALL (FIRST HOT EXPERIMENT) UHDERPREDICT!OH OF POMER IH PINS HEAR !IATERHOLES - ONLY PART RECOVERED BY t10DEL CHANGES (SEE CEHPD"l53)
BUT MAS COVERED BY Uf'iCERTAIHTY.
1977
-FIRST DIT (t1ULTI-GROUP TRANSPORT THEORY) CALCULATIOHS COt<PLETED-
'=-'PREDICTIOr< OF WATERHOLE PEAKIHG QUANTIFIED.
-LEAD TO CURRENT ACTIOiH TO ACCOUNT FOR EFFECT.
10
CAL ATIOH OF DES IGiH PO! JER PEAKS CRITICAL LATTICE EXPERIMENTS C
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UHCERTAINTY AND BEST METHOD BIAS DIT CALCULATION DES IGH METHOD BIAS AND UHCERTAIHTY REACTOR LATTICE PDQ-DIT DIFFERENCE DESIGN MODEL CALCULATIOiH PIH PEAK POi~iER PD<>
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I q
V lc
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~ IN 14 x
14 CO
- UEL, Fp BlAS I ~
N
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(
-1 ) x 100%
PDQ
~ N
~
g l,.<<,c 10
. 14 16 18 20
'ROCEDURE FOR THE CALCULATION OF PIN PEAKS 0 000 0
ASSEMBLY GEOMETRY CORE GEOMETRY DIT (INTEGRALTRANSPORT)
PEA~A.E IFP)
STD. DESIGN PDQ (CEPAK 2.3)
PEAK/AVE (FP)
FP BIAS'FOR OPERATING CORES STD. DESIGN PDQ (CP2.3)
DIT vs CRIT.
EXPERIMENTS BlAS PIN PEAK AND INSTRUMENTW'
APPROXIMATIONS TN DET
+ PIN-CELLS COUPLED BY IiNTERFACE CURREiNTS CONSTANT OYER EACH FACE
- DOUBLE Po AT EACH INTERF/CE
{SEPARATE YERSION MITH HiGHER ORDER TERf1S IN EXISTENCE)
- COLLISION PROBABILITIFS WITHIN EACH CELL.CALCULATED FOR CYLINDRICIZED 6 EGA/ETRY (OPTIONAL:
SQUARE BOUiNDARIES IN CALCULATION OF CP'S
e C R I' C A L E X P E 8 I M I Vl I T H C E A bV A T E R I< 0 L E S
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EXPERIV.ENT Eld R. 'i'o V /V(
T PP,".I SIZE OF CORE No, OF PINS YEAS, AT 'V'H.
No, OF IVATERIiOLES 14 x 14 16x 16 3 vr/o 3 vr/o 1.63 1.71 4 ~ ~
V%
J
0 'ALVER lFFS l
I I
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~I lNF 1 NlT'E LATTlCE OF C'1 2 LO ASSEM8LY THERMALFLUXx 10 2 DEVlATION(%} FROM HORSE
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1 h'AHORSE, HET, 2
NUTEST - D17, K=1, HET..
3 DOT, S8, 28 x 28, HOM.
p
~
mr
~
~
Cl
0
~
(o/r,<->) x rood.
D
=
OIT M
~
bAEAS/G LO8AL
~wu"
. KR ITZ UVlH2
r 0
DIFFERENCES FRO)1 CHAPTER 4 OF CEHPD-153.
ENLARGED DATA BASE INCLUDING C-E CRITICALS PEAK PIHS SPECIFICALLY ADDRESSED.
CEHPD-153 GAVE:
)(
=[ ](CALC. BIAS) ka =t j (CALC. UNCERTAINTY)
DIT vs.
Exo.
OBSERYED BIAS AND STANDARD DEYIATION FOR PINS FACING CEA >JATERHOLES LATTICE HO.
OF POINTS TOTAL tlEAS.
BIAS S
'STD.
DEY.
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SECTION 2 Actions to Accommodate Increased t!aterhole Peakinq for St. Lucie Unit 1
(
The purpose of the information presented in this section is to demonstrate that the analysis package used for the first cycle of the subject plant to date has employed methodology with sufficient conservatisms to accommodate the increased water hole peaking discussed elsewhere in Section 1.
The only Technical Specification changes required are a tightening by 0.5/ of the surveillance requirements on in-core or ex-core surveillance of peak linear heat rate (kw/ft) relative to the LCO.
Since the subject plant has been operated with at least 0.5/ margin to the operating limits, there have been no violations of these limits (Slides 1 and 2).
This summary will take the form of a cormentary on the remainder of the slides presented to the NRC Staff on December 16, 1977.
Slide 3
displays the sources of margin which offset the increased water hole peaking for Calvert Cliffs Unit 1 Cycle 2, as an. example.
The top half of the slide deals with DHBR-related limits - the LSSS and the LCO.
The lower half of the slide deals with kw/ft-related limits-the LSSS and the LCO.
Let us consider first the DHBR LSSS.
The uncertainty allowance employed in the analysis whose resultant setpoints are contained in the submittal for Cycle 2 of the subject unit was 5A.
- However, the total allowance which must be made is 9.8Ã, of which 4.6/ is the increase in water hole
- peaking, and 5.2X is the INCA measurement
a r
uncertainty on Fr.
It, will be remembered that INCA is used to surveille Fr relative to its limit.
The technical specification
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limit on Fr is an assumption in the setpoint analysis.
INCA uncertainty-must-also be.accounted for in the setpoint analysis, since it does not appear explicitly in the surveillance requirement.
The next column of Slide 3 shows that. an additional penalty of 4.8% exists (9.8: minus 5%).
The margin credits which offset the additional penalty are
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displayed in the column headed "REMEDIES".
There is a 3% credit.
for improved thermal margin/low pressure trip synthesis.
This was summarized on Slides 4-7 of the December 16, 1977 presentation, but will not be discussed further here since a detailed presentaticn of this topic appears in Section 3.
A 2% credit arises from the statistical treatment of selected uncertainties. 'his topic is treate'd in detail under the heading for Slides 7 and 8.
Since the two listed credits offset, the 4.8%
penalty there is no requirement. for changes to the technical specifications or the setpoints in connection with the DNBR LSSS.
The next line of Slide 3 deals with the DNBR LCO.
The same additional penalty of 4.8% margin must be accommodated as for the DNBR LSSS.
3% credit is taken for margin gains computed by TORC ir-conjunction with the CE-1 correlation relative to the design basis COSMO/INTHERMIC analyses employed for these operating reactors-The basis for the 3% credit was summarized on Slides 9 through ll in the December 16, 1977 meeting, and will not be dealt with
further here since Section 4 contains a detailed discussion of this material.
The remainder of the 4. 8%, additional penalty on DNBR LCO is'ffset, by the credit for the statistical uncertairity treatment covered under the heading for Slides 7 and 8.
Ne now invite attention to the line of Slide 3 headed "K)i/FT LSSS".
The setpoint analysis for the kw/ft LSSS contains a
10% uncertainty allowance at present.
This must be revised to 10.4% to take account of the 4.6% increase in wate.". hole peakin" and the 5.8%
XNCA allowance on Fq measurement.
There is, thus, a
0 4Q additional penalty to be accommodated.for the kw/ft LSSS (10.4% minus 10%).
Xn the "REMEDIES" column we see that this penalty is offset by a credit from the sta"'istical uncertainty treatment (Slide
- 8).
Therefore, no change is required to the technical specifications.or to the setpoints..
The kw/ft operating limit monitoring -process presently includes an 8% allowance foz uncertainties.
- However, a 10.4$
allowance must be accommodated, leaving an additional penalty of 2.45.
Of this, 1.9% is offset by the statistical uncertainty treatment (Slide 8),
and the balance by a tightening of the relevant portions of the technical specifications on kw/ft surveillance by 0.5%.
e For reference, some data is included at the bottom of the
,t slide on 'variations of Fr and Fxy measured at the time in life when these quantities were their highest during steady state operation and the technical specification limits on these quantities.
Slide 7 displays the source of the 2% credit on the Fr
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uncertainties arising from the use of statistics to combine se-lected j.ndividual uncertainties.
Xn the rod bow topical (CENPD-225-P) was presented a computation of Fr uncertainty by statistical means in comparison with the multiplicative combination of the individual uncertainties.
The terms which enter into the uncertainties on Fr include the engineering factor on Fr, the nuclear uncertainty on Fr, and the fuel rod bowing factor.
(I) of Slide 7 reproduces the computation precisely as presented in the topical report for reference, with the conclusion that the employment of a statistical combination of the uncertainty components discussed in that report yielded an overall uncertainty 2.35-lower than just the multiplicativc.
combination of the engineering and nuclear factors.
The setpoint analyses, which form the basescof the submittals for the initial cycles and reloads for. the plants in question; included a multiplicative combination of the nuclear and engineering components of the overall uncertainty.
We may take credit for the statistical combination of the relevant uncertainties only after using the app opriate numerical values for the components, and showing that these components meet the normal criteria of randomness and independence.
The arguments for randomness and independence for these components are presented in detail in 0
CENPD-225-P and will not be repeated here.
II of Slide 7 shows the computation of the. uncertainty credit using the appropriate value of-5.2X for the Fr uncertainty with a 2"result,
~
~
~
~
~
~
including the contribution of the fuel rod bow factor.
Slide 8 is the analog of Slide 7, but for the Fq uncertainty.
By a similar process we compute a 1.95 credit for the use of a statistical error combination including the fuel and poison rod bowing factors, relative to the multiplicative combination of just the engineering and nuclear factors.
It should be noted that the random component of the INCA uncertainty on Fq is 5.7X, and that X 'is O.l/ for a total INCA uncertainty on Fq of 5.8/ as seen on Slide 3.
Slide 12 summarizes the situation for Cycle 1 of St. Lucie-l, which is expected to be completed about April 1, 1978.
Uncertainty allowances of 10/ and 85 were included in the setpoint analyses for the DNBR LSSS and LCO, respectively, with the-result that the only addit,ional penalty required was 1.8/ on the DNBR LCO.
This is entirely offset by the qredit from the statistical uncertainty treatment.
The kw/ft situation is precisely the same as for Calvert Cli;"fs-1 Cycle 2 discussed above.
Slide 13 summarizes the Technical Specification changes required for implementing the 0.5Ã tightening of the surveillance on peak linear heat rate.
In conclusion (Slide 14), existing margins and conservatisms in the setpoint analyses for submittals now in the hands. of the NRC Staff compensate for the increased water hole peaking, with the exception of a 0.5X tightening of Technical Specification surveillance requirements
" r peak linear heat rate,
STATUS OF OPERATIt,"6 C-E PLANTS"AND PROPOSED ACTIOff TOPICS PLANT-AilD-CYCLE-SPECIFIC EYALUATION 1'MR6IN CREDITS If'.PROVED Tl'~fLP TRIP SYNTHESIS
-TORC/CE-1 CREDIT FOR DNB/LCO STATISTICAL UNCERTAINTY TREATffEiiT SUYifWRY
le NO CHARGE IS REQUIRED TO FR, FXy TECH SPEC
~
LItiITSg BECAUSE l I
2s THE INCREASED PEAKING IS ACCOf10DATED BY EXISTING MARGINS AND CONSERYATI SMS, 3 s TECH i SPEC SURYE I LLANCE REQUIREMENTS MUST BE TIGHTENED BY 0 s 5'OR I N CORE OR EX CORE KN/FT LC0 s
0c SINCE THERE HAS BEEN ~~0,5R MARGIN BETNEEN ACTUAL OPERATION AND THE OPERATING LIMITS ON KN/FTi THERE HAYE BEEN NO VIOLATIONS OF THESE LIf.lITSs
4
CALVERI CLIFFS UNIT 1, CYCLE 2 EFFECTS ON LSS LCO:
(NOH OPERATING UTS DO>lN 1/78)
R.
ED PRESENT FR
- UNCERT, ALLOW, REVISED FR ADDITIONAL
Af'lT, CHANGES TO
- TECH, SPECS, 3R'LSSS 5%
Ll,6 + 5.2
= 9,8 Ll.8%
1,'3MPROVED TM/LP l R I P SYNTIlES I S 2alCREDIT FROM STATISTICAL UNCERTAINTY TREATMENT 3%
2/i l'!ONE 3R LCO 5%
4,6 + 5,2
= 9,8 0.8%
1 CREDIT FRQM TORC/CE-1 2,; CRED I T FROM STATISTICAL
'JNc E RTAI NTY TR EATr~ENT 30/
2%
PRESENT FR/FxY UiHCERT, ALLOW, REVISED F IF ADDITIONAL Uf')CERT, ALLOW.
PENALTY REMEDIES TYPE ANT.
CHANGES TO lQ']/F'T TECH.
SPECS
'FT LSSS 10%
0.6 + 5,8
= 10,9 O.Ll%
1,
- CREDIT, FROM STATISTICAL
.UNCERTAINTY TREATMENT 0, LlX
-HONE-
~FT LCO tf
= 10.l}
3.s IN-CORE I'lONITORING REBUCE ALABf". LIMITS 2,
EX-CORE HONITORING CHANGE SCALING EQUATION CRED IT FROM STATISTICAL UNCERTAINTY 0,5%
1.9%
YES CURRENT r'VVGIr'>S Or':
QUANTITY TEGRATED RAD IAL-'.(FR'):
'OAR R<A?AL MEASURED (3,970 Nl'JD/T),
1,35 l.0t TECH.
SPEC.
LIMIT l,tl3 l.5O ERG I H 5.9%
-. 6.47m
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560
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550
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540
~~530
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520 C) 510 500 50 70 80 90 CORE PONfER (% GF RATED'10
~~
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Il
'5 3
8-m 560
~
550 5(0
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520 C) 5l0 500 50 70 80 CORE POKIER (% OF RATED)
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FR UNCERTAINTY U
s COf1PUTATIOH OF F~
IN ROD BON TOPICAL REPORT, (CEHPD 225P)
E N
F A
RSS COHB IHATIOH OF FR'R'R F =1+
U E
2 N
2 F
2 (F
1)
+ (F -1)
+ (F -1)
R R
R E
FORt FR = ENGINEERI'HG FACTOR
= 1,03 N
FR = NUCLEAR FACTOR
= 1,06 F
FR = FUEL ROD BONING FACTOR = 1s0" 80 E
N Bs HULTIPLICATIVE COYiBINATION OF FR AND FR F
= (FR)
(FR) = 1.0913 U
N I I s'OHPUTATIOH OF FR F
R FR 1
s 052 E
N F
As RSS COMB I HATION 'OF FR/
FR~
FR U
F
= 1,0627 R
E N
B NULTIPLICATIVE COtSINATION OF FR AND FR U
FR = 1,0836
~
~
1.0856 - 1,0627
= 0,0209:
2Z
'I0 Cs CREDIT FOR STATISTICAL LIHCERTAINTY TREATMENT:
F UNCERTAINTY h
U CoMPUTATION QF FQ IN RoD Bor( ToPIGAL RFPORT
((;Ei'lPD-225P)
A; E.
N F
P RSS COYBIHATIOH OF FQ>
FQg FQ~
Fq U
E 2
H 2
F 2
P 2
FQ
=
1 +
(FQ-1)
+ (FQ-1)
+ (FQ-1)
+ (FQ-1)
FORl 1.0898 E
FQ EHG I iHEER I HG FACTOR N
FQ = NUCLEAR FACTOR
= 1.05 F
FQ = FUEL ROD BONINC FACTOR
= le018 P
FQ POS I OH ROD BOW I NG FACTOR = 1, 021 E
'N B
YULTIPLICATIVEColiBINATION OF FQ FQ FQ = (FQ)
(FQ) = 1,112~
U N
I I c COMPUTATIOH OF FQ FOR FQ 1
g 057 E
N A,
RSS COMBIHATIOH OF FQJ FQJ F
= 1,0701 Q
F P
FQ, FQ E
N B
HULTIPLICATIVE COMBINATION OF FQ AHD FQ FQ = 1,0887 CI CREDIT FOR STATISTICAL UNCERTAINTY TREATMENT 1.0387 - 1,0701 = 0,0186 1 9K
O C
~
~
f OVERPO'i'lER MRG I N GAIN FOR DNBR LCO HIGHER
~ t
.TORC/CE-j.
ERPO')IER GIN PARTIAL TORC CREDIT (3X)
'OSY<0/<(-3 LOHER
TORCICE-1 GAIN NUMBER OF POINTS IN PARAMETER SPACE AVERAGE GAINS
% OVERPONER OPERATING POINT 6AINr K 0VERPOHER
I USE OF TORC/CE-1 CREDIT COHCLUS I Ot'1 r
II PARTIAL CREDIT: FOR TORC/CE-1 H!LL BE TAKEH IN LCO (3/o)
LOFA IS 1IVITING AOO.FOR BoTH COSi]0/kt-3 PHD TORC/CE-1 I
REQUIRED f'4RGIH DOES NOT CHANGE PROTECTION SYSTEH CONSTANTS DO NOT CHANGE INITIAL COHDITIOHS ASSUMED IN SAFETY ANALYSIS ARE HOT EXCEEDED IH PLAHT OPERATIONs r
h 0
l
CO:
PRESENT FR UiNCERT
- ALLOM, RE/ISED PR
('lOl'f OPERATI'l6, SHO Ot'i3
- /78) i
'E('
mal C'-'P'O'S i-"
- TECH, SPECS, i LSSS 10/ ~
0,6 + 5,2
= 9,8 NOT REauIREO tlONE l LCO 0,6 + 5,2 9
n 1.
CREDI:r FRoi<<lrATISTICAt, 2Z UNCERTAINTY lREATWENT ilONE PRESENT F /FXY U lCERT, ALLOi'I, REY ISED FR/FXY Ut'~iCERT,
- ALLOW, ADDITIOiiAL PEtNLTY
.REi'iEDIES CHAf'AGES.TO KH/zT TECH, SPECS;
=i LSSS 3;0%
9,6 + 5,8
= lo.~l O,tl/
1, CREDIT FRON STATISTICAI
.UNCERTAINTY TRLATYENT 0,0%
~ ttlO Nic ~
=T LCO CUR".=MT i'NRG INS Oil:
Q,6+ 5,8
= 10,~>
2 3,
IN-CORE I'ION ITOR ING REDIiCE ALARl'! LIMITS 2 i EX-CORE i"lONITORI NG
( HANGE SCALING EQUATION 3,
CREDIT FROi~i STATISTICAL tjNCERTAINTY 0,5%
1,9%
YES i"i'if<i!T t TY NEASURED (7380 YhiD/T) i,'
TEC!3,
- SPEC, LINIT i'Ni'iiI R
[
ECRAT O RADIAL (F )
1)32
'i ZPi 1,36 1.3R 3,0K 1.RR
TEGi SPEC QW$iES Ii&COPZ tOJITNIffS ADD PENALTY FACTOR OF lsK6 TO EXISTING PENALTIES USED TO SET IN-CORE ALARI'l LINITS EX-COPZ fOtITORIf'lG
'-1" fIhDIFY EX-CORE SCALING EQUATION TO CONTAIN A FACTOR OF 1,N5 TECH SFrC SECTIG'J AFFECTED SURYEILLA'iCE REQUIRE "NTS PORTION OF LINEAR HEAT RATE TECH SPECl ST a LUCIE-1 SECTION 5/0,2,1 IF
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j.s NO CHAi";GE IS REQUIR D 70 F
g F
TECH'PEC R'Y LIWITS, BECAUSE:
~pv
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2 THE INCREASED PEAKIHG I S ACCOf"iODATED BY EXISTIf'lG f'"RGIHS AHD CO"fSERVATISf)S.
TECH SPEC I SURVEILLAiHCE REQUIREMENTS MUST BE TIGHTEHED BY 0,5'OR IiH-CORE OR EX-CORE KNlFT 1I.Oe I
0e SINCE THERE HAS BEEN OiSZ MRGIN BETi'CREEN ACTUAL OPERATION At(D THE OPERATING LIHI'TS ON KNIFTg THERE HAVE BEEH fiO VIOLATIOHS OF THESE LIf~iITSs
.. SECTXON 3
XNPROVED TiiERt'RL IWRGXN/LO'i0 PRESSURE TRXP SVNT/)ESXS The purpose of this section. is to describe the improved Thermal tlargin/Low Pressure (TN/LP) trip synthesis methodology for which a 3%, credit is claimed as discussed in Section 2-The generation of data, as described in CENPD-199-P "Topical Report on Setpoint Methodology," remains unchanged.
The change, d'-
~
for the thermal margin limit lines.
The generation of the thermal margin limit lines and the Pfd vs.
X curves remains as describea fdn p
in CEVPD-199-P.
Slide 1, which is Figure g-18 from CENPD-199-P, is a summary of the previously used TN/LP LSSS synthesis methodology.
A.brief discussion of this slide is included here so that a base is provided lfrom which the changes, resulting the improved synthesis methodology, can be understood.
3, 5
m et¹¹¹¹A4 ¹¹, ¹;
3 6
-The combined results of these two procedures is an equation c
the form shown in Block F which is the T?1/LP LSSS de ined in term of pressure, axial shape index, power and inlet temperature.
Slide l is then a summary of how the thermal margin limit lines and Pfd vs. I curves were previously combined to synthesize fdn p
the TH/LP LSSS.
Slide 2 is Slide l modified to show a summary of the improvec synthesis methodology.
3, 5
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M~~W~~I I)i y~$f'I ll*
t'
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t $%
t 1
1'd 13 'llustrates the magnitude of the improvement, resulting fro.. use of the improved T. /
'p y
~
+
~
'I/LP tri s nthesis methodolo~.
This slide shows the results of TH/LP LSSS calculate.ons for a 2700 Nwt NSSS using both the orzgxnal and the imp
~
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e im roved synthesis methodology.
T e h
Tfi/L> LSSS from this slide is repeated.
on Slide 14 for comparison with TM/LP LSSS generated using the improved synthesis methodology but assuming a
p radial eak 3%
higher than that used for SliQe 13.
Slide 14 shows that, even with a 3% higher radial peak, the improved methodology results in an improvement over the original synthesis methodology.
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TM/LP DN 8 LS S SYNTHESIS
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DMS OVEH,PoQER.
VRBTRVzoN AZYH caRz.
- PoazR, R,C FCA.ENC.E.
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0 SECTION 4
PARTIAL CREDIT for TORC/CE-1 in ASI l1ONITORING LIMITS SLIDE 1 The purpose of this portion of the slide description is to explain the technical details of the plan to take partial credit for TQRC/CE-1 thermal margin gains in the monitoring system, i.e.f the ASI Monitoring Limits.
No credit is taken for TORC in the trip setting.
Basically, the objective is to demonstrate the existence of inherent conservatism in the "open hot channel" design code COSY~0/N-3 relative to the "open core" design code TORC/CE-l.
Some backg ound on these methodologies will provide the thermal-hydraulic bases for the conservatisms.
There are three fundamental models used in the thermal-hydraulic analysis of pressurized water reactors.
The closed hot channel, the open hot channel, and the open core methods.
Closed hot' channel postulates coolant in a sub-channel bounded by four (4) fuel rods, is heated through its axial flow upward
'0 through the core without any mass or'energy exchange with its neighboring sub-channels.
Xt can be viewed as a sub-channel with an inpenetrable lateral boundary.
The axial pressure drop across the core computed using the'core average flow rate,'provides the boundary conditions for determining the closed hot channel flow rate.
That is, the flow rate in the closed hot channel is
raised or lowered as required until the axial pressure drop in the closed hot channel equals the core average pressure drop.
W The COSMO code is a open hot channel thermal hydraulic model.
That is, it permits an exchange of energy from the hot channel with its neighboring sub-channel.
The cpge
" does not solve the energy and momentum equations explicity.
- Instead, separate analysis is performed to determine mixing factors.
Mixing factors are defined as the enthalpy rise in a axial length of the hot channel where turbulent interchange of coolant is allowed divided by'he enthalpy rise in the same length of hot channel where no turbulent interchange is allowed (e.g.,
the closed channel enthalpy rise).
Mixing factors are numerical values less than unity.
They -are applied as multipliers on the hot channel enthalpy rise.
Another feature of the COSMO model is in the evaluation of the hot channel flow rate.
The boundary conditions start with the average pressure
- drop, but p'roceed to the hot assembly to determine the local nodal axial pressure gradient.
imposed on the hot sub-channel.
Important features of the COSMO open hot channel analysis are that it computes the impact of turbulent interchange of coolant in the hot sub-channel with its neighbor sub-channels and it places the hot. sub-channels in the hot assembly for the determination of axial boundary conditions for the 'computation of sub-channel flow.
The TORC/CE-1 code was developed from 'the COBRA-IXI code. described
4
in the literature.
The TOBC/CE-1 code differs from both of the
~
~
models described above in that explicit solution of the momentum and energy equations is performed on a core-wide basis, Both turbulent interchange (mixing of coolant with the effect of reduced enthalpy with no net mass exchange) and divergent cross flow (the net exchange of mass from one region of the core to another) is permitted if calculated to occur.
SLIDE 2
It is important to note that the thermal-hydraulic design bases 0f the plant does not change.
COSHO/Ã-3 remains the model on t
which the safety analysis is performed.
No credit is taken in the trip settings for the inherent conservatisms in the COSh!0/N-3 irodel relative to TORC/CE-1.
COSNO/N is used.to identify the limiting anticipated opexational occurrence and in the: assessment of the required margin.
The acceptability of the consequences of the analyzed events is based on the DNBR computed using COSIIO/N-3.
Therefore, the transient sa ety analysis does not change and need not. be repeated.
The required margin maintained in the monitoring system as. quantified using,COSHO/N-3 does not change and no credit is taken in the thermal margin low pressure limiting safety systems settings.
STRIDE 3
This slide shows the type of margin gain which we have identified by taking analysis of hundreds of cases where process variables
are permuted.and running these cases with COSMO/h'-3 and
~
~
TORC/CE-l.
The Y axis shows overpower margin.
The overpower margin is defined as the percent or fraction of rated core power at which a DNBR limit is achieved in the analysis.
The margin gain shown as the difference in margin as indicative of the inherent conservatism in COSMO/Ã-3 relative to TORC/CE-l.
'SLIDE 4
'i This slide shows explicit analyses of the comparison of COSMO and TORC for several hundred
- cases, This analysis simulates the 3
5 Pfdn analysis-Here using
. typical shapes, there is at minimum thermal margin gain by using TORC relative to COSMO.
The maximum gain in thermal margin as evaluated from this data equal's 3,
5 about SLIDE 5 For an analysis over all operating
- space, 691 data points were examined.
This data was representative of Combustion Engineering's 16x16 fuel assembly, Operating space in temperature,
- pressure, flow, integrated radial peak and axial shape were permutted.
Results of this indicate an average. margin gain of 7%-overpower margin.
As was stated on the previous slide, the operating point margin gain eauals about 12% overpower.
From this analysis, Combustion Fngineering will claim a 3% net credit, for TORC in auantifying the inherent conservatism in the open hot channel COS?50/N-3 thermal margin code.
Although the data was generated for the 16xl6 roactor fuel,
4
t there are no fundamental thermal-hydraulic concerns with regard to the outcome of an analysis of Combustion Engineering 14xl4 fuel using a similar data base.
That is, we anticipate the thermal margin gain, if explicit analyses on 14x14 fuel were to be performed, would be of similar magnitude.
This 3-o is applied in the ASX Ilonitoring System.
I I
SLIDE 6
The conclusions from this analysis
- then, are that a
3-o TORC/CE-1 limiting condition operation credit is a conservative estimate of the increased available margin relative to COSNO/Ã-3.
A typical 16x16 Pf analysis has shown minimum and
. maximum gain.
fdn Over all operating space a
7S average gain has been demonstrated.
This latter assessment was performed over a wider parameter space than allowed by current 14xl4 plant, technical specifications.
SLIDE 1 THE PURPOSE OF THIS PORTION OF THE YiEETING IS TO EXPLAIN THE TECHNICAL DFTAILS OF THE PLAN TO TAKE PARTIAL CREDIT FOR TORC/'EF-L THER."iAL hARGIN GAINS IN THE Y!ONITORING SYSTEMS (ASI HONITORING LIMITS) a
SLXDE 2
~
~
e
MAJOR FL~E"i NTS le DESIGN BASIS DOES NOT CHANGE COSf~0/1".-3 IS USED AS THE DESIGN BA'SIS UPON "NHICH THE SAFETY ANALYSI S IS PERFORMED'.
LIMITING AOO Ba QUANTIFIES THE REQUIRED f4RGIN C,
ACCEPTABt E CONSEQUENCES
=
4 g THEREFORE!
TRANSIFNT ANALYSIS DOES NOT CHANGE<
REQUIRED MARGIN MAINTAINED IN THE MONITORING SYSTEM DOES NOT CHANGE, HO CREDIT TAKEN IN TN/LP LSSSi
OY R
0>>
TOPiC/CE-1
~
~
OYERPONER RGIH VSs ASl Les ego
+"'COSi~i0 e
TORC UJ CO C
UJ (D
AXIAL SHAPF Ii~(DEX
L'
SLIDE 5
~
~ ~
~ Q Q TORC/CE-1 GA IH Ih"iBER OF POIHTS IH PARQ'iETER SPACE 691 AYEPAGE CAIN K OYERPO ER OPERATING POINT 6AIH K OYERPOh'ER
c
SLIDE 6 COi'IC~U~S10>>
5% TORC/t:E-t LCO CREDIT IS A CONSERVATIVE ESTIHATE OF THE INCREASED AVAILABLEHARGIN RELATIVE TO COS,"'10/t'l-5.
A TYPICAL ~6 X 16 PLANT ANALYSIS HAS SHO)'lN YiINIHUH AND HAXIl)UH GAINS'HIS ASSESSMENT NAS PERFORNED OVER HIDER PARAMETER SPACE THEN ALLONED BY CURRENT 10 X 10.PLANT TECHs SPECS'
~ Cl L~