ML20040H221
| ML20040H221 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/19/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170457 | |
| Download: ML20040H221 (25) | |
Text
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{ TOPICAL REPORT EVALUATION
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ENCLOSURE 2 Report No.: WCAP-8567 Report
Title:
Improved Thermal Design Procedure Report Date: July,.1975 Originating Organization: Westinghouse Summary of Topical Reoort 5
The overall objective of the thermal-hydraulic design of the reactor core is to provide adequate heat transfer such thab:
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1.
Fuel damage is not expected during normal operation, operational transients (Condition I), or transient conditions arising from faults of moderate frequency (Condition II).
2.
The reactor can be brought to a. safe state following a Condition III event (acciden3.'~1.th.o$1y a$ sEI11.frNtion of.. fuel bds daked,
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w 3.
The reactor can be brought to a safe state and the core can be kept suberitical with acceptable heat transfer geometry following transients arising from Condition IV events (accidents).
One design basis for Condition I and II events is protection from departure from nucleate boiling (DNB). The subject topical report describes a new approach to the DNB design basis for Westinghouse PWRs.
The proposed DNB design is based on consideration of uncertainties in plant operating parameters, fabrication parameters, nuclear and thermal parameters,.
and the use of the, appropriate DNS correlation for the plant.
The proposed design basis is:
there must be at least a.95% probability tnat the minimum DNBR of the limiting powe'r rod is greater than or equal to the DNBR limit of the correlation being used.
Parameter uncertainties or variances obtained 3:9 :ne evalut:i:n :f din 1 e tetsmir.ed at a 951 c:nfidsnce levei 8202170457 820212 PDR ADOCK 05000445 O
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2-The current design procedure is to show.,e.g.. that,the W-3 DNBR > 1.30 on the I
limiting power rod considering a,1;l parameters at fixed,gonservative values (i.e.,'
transients were initiated from conservative conditions ~with fixed offset in power, TlN, Fyf, pressure and flow from nominal values).
The proposed. design procedure is based on the statistical combination of the effects on DNBR of uncertainties in vessel coolant flow, core power, coolant inlet temperature, system pressure, effec'tive core flow fraction, N
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F g (rod), F g (assembly) and F g, The. nominal value for each of these parameters is used in the analysis.and the effect of the uncertainties on DNBR is added to the 95/95 limit for the correlation. Other design factors N
,such as TDC and F are taken as fixed conservative values.
In order to relate the variations in design parameters to DNBR variations for the proposed method, an uncertainty factor, def,ined by the following, equation, is used:
y=DNBR(variable)/DNBR(nominal).
(1)
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The value of DNBR (nominal) is detemined by considering the values of all the design parameters to be at their nominal or best estimate values. The value of DNBR (variable) is based on values of the d'esign parameters including O
y
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their uncertainties and deviations from nominal values.
The DNBR uncer-tainty factor is assumed to be affected by changes in the values of the
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' design parameters according to a relation of the form:
dy, 1 n
i dX y
S i
g (2) 7=1 X g i
The value of S can be interpreted as representing the percentage change g
in DNBR resulting from a one percent change in X. all other parameters being held constant. For small perturbations from mean values of X, for Xg independent, and for constant values of S, equation (3) can be obtained.
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s
[')h.:
g $'2[-O )L T
i-g
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(3)
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where:
((=meanvalueofX,
[= variance of X,
vp = coefficient of variation.
In order to determine the sensitivity factors, the standard THINC-IV thermal design methodII) for determining minimum DNBRs was followed. A reference case was established by setting all input parameters to their nominal values.
Each of the parameters considered in the DNB thermal design procedure was then changed in value o'ne at a time and the resulting DNBR for the peak power rod The sensitivity parameters, S, wire obtained from plots of DNSR versus noted.
individual parameters.
Input parameters were reset to values other than the reference case remir.11 ni.es so that a wide range of DNBR values c:uld be covered.
Cbserved variatiens in the 5, were generally small over a wide 9
range of conditions.
In all cases, the largest numerical value of the 1
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sensitivity factor over the DNS range of interest was chosen'for'use in the DNB analysis. Ranges of parameters for which the sensitivity factors are applicable are given in Tab 1'e 1.
The nominal value and variance of each input parameter is calculated for plant operating parameters, nuclear and thermal parameters, and fabrication parameters.
Discussion of the method.fer obtaining variances 'for two plant parameters should suffice to illustrate the overall procedure.
Ba, measuring the steam generator power outpu't and the coolant temperature V
rise, the coolant flow in each loop of a reactor can be determined. The results of a comparison of measured and predicted flows for six Westinghouse reactors are shown in Figure 1.
The percent difference in predicted and measured flow for this set of data leads to a ' standard deviation of 0.027.
A 95 percent upper confidence value for the population standard deviation is 0.0377.
Considering this total variation to be composed of additive-
. uncertainty and accuracy components leads to the relation:
I/t U~
Y u
(4) where the ' subscripts t, U and a refer to total, uncertainty, and accuracy, respectively.
The accuracy uncertainty is estimated to be equivalent to a value of 7, of one percent.
Substitution of this value into equation (a) c yields:
v = [(.0377)2 - (.01)23 U 2 =.0353 (5)
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its the nlue used in :ne :nU-al design analysis for the coclant flow stancard deviation.
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dde to cal.orimetric core pcwer is consider;d pressure A maximum error 12 percent in flow, feedwater temperature, s errors in measuring feedwater If the*kncertainty in core power 2 ion carryover.
- 5) (4) = 1.8 i
and moisture be uniformly distributed over The 95 percent number is 2-(.0 the 1
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of 2/ h = 1.2 p'ercent results.
f i n I and II DNBR" applicable to all Condit od va percent.
i The minimum allowable or " lim tthe D events is determined by usingThe method. for limiting power parameters. a ONB correlation which pred c5' percent From the Figure 2 for f 1.30.
95 percent probability, with 9leate boiling at a DNB ratio o rod will not depart from nuc n uncertainty factor at thei ed from norm mean and standard deviation, ad us d by the bility level is obtaine The 1.3 value from the bility tables.
is less than 1) to obtain the l m uncertainty factor (which ditions.
that transient at best estimate design conII event the minimum DNSR dur ng i
In evaluating a Condition I'or All input parameters are the V procedure.
flow fraction or is calculated with the THINC-Ifew, such as. hot assembly Input para-lues.
i nt, which use limit or fixed v nominal values except for a ted are thermal diffusion coeffic e minimum DNSR is associated with the transientl es fo meters adjusted to appropriate va u calculated.
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With the new thermal design procedure, uncertainties "n protection system setpoints may also be treated statistically.
For example, for the over
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.s temperature AT trip, instrument errors may be combined into subgroups of independent uncer,tainties and the subgroup errors combined statistically.
This is done by assuming that each subgroup uncertainty has a uniform i.
probability density ' function ~ between the error limits, calculating the standard deviation for each subgroup and taking the root mean square of
,all the subgroup standard deviations to give the total standard deviation.
The total uncertainty in the overtemperat.ure aT setpoint at a 95 percent probability level is then 1.645, times the total standard deviation.
This to' tai uncertainty is subtracted frcm the maximum allowable trip setpoint to yield the nominal setpoint.
The ONB core limits are the locus of points in core power, inlet temperature and pressure along which the minimum DNBR is the limit value. The core limits are calculated by using the THINC-IV computer code. All parameters are assigned values (primarily nominal) except for iniet temperature, pcwer and pressure which are varied to determine the points at which the limit DNBR is obtained.
Staff Evaluation The traditional method of protecting against departure from nucleate s'
boiling (DNB) in PWRs,' referred to by Westinghouse as the fixed value method, provides a conservative estimate of the minimum DNB ratio.
The fixed va he e: hod established 1 tinid.m 2.ER limi, e.g., 1.30, based on the 95/g5 criterion given in the standard review plan; parameters such as F ad core 4 pass new are taken as Nxed conserva:he vahes and g
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initial conditions for transient analyses such as power, pressure and I
inlet temperature are offset by,. fixed, conservative amounts. The fixed value me'thod is acceptable as a licensing method but it includes a large conservatism rela,tive to best estimate conditions.
The proposed DNB design procedure removes a large portion of the conser-vatism inherent in the fixed value method. The proposed procedure results in a DNBR benefit of approximately l'O to 15 percent as applied to D.C. Cook Unit 2, and even larger benefit if full credit is taken for the method as described in WCAP-8567.
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Because this large reduction in DNBR margin results'from a statistical
' combination of uncertainties, the parameters treated by the statistical combination were examined to assure that all uncertainties had been considered either statistically or in a bounding fashion, and that those, parameters treated statistically were treated correctly.
Computer code uncertainties were not considered extensively in the topical.
report but were discussed at length in responsh to staff questions. Uncer-tainties in the THINC-IV code and the LOFTRAN code were investigated by the staff because LOFTRAN is used to model plant transients and THINC-IV is used to calculate fluid conditions and DNBR.
~ The staff believ.es that. conservatism exisJin tha mthed of plant protection, the Nethod of analysis and code input to offset code uncertainties but the conservatism has net been quantified by Westinghouse.
.Neitner nas the code uncer ain;y :een qua..:ified.
Wes:ingneuss atts ::st ::
quantify THINC-IV uncertainties but'the reported uncertainty is consideraoly I
less -han ex;:erience with other similar codes, such as CCBFA, would indica e.
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The input parameters shown in Table 2'were presented as evidence ~of conservative inputs for the transient codes and the SAR values are definitely A
. conserva,tive relative to the expected values; however, most of the SAR values are given as allowable values in plant Technical Specifications. Although it is unlikely that all the parameters will be at the Technical Specification value simultaneously,. one or more may be at its Technical Speci.fication limit at the same time.
Therefore, use of the Technical Specification values in safety analyses results in an unquantified conservatism.
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Uncertainties in transient codes will generally have only a minor effect on the minimum DNBR for any of the transients analyzed by the new design pro-cedure. This is because the reactor trip is based on measured values of pressure, TAM, aT, flow, neutron flux.and low voltage (to the pumps).
Transient events analyzed with the proposed thermal design procedure may be grouped into those transients within the design basis, of.the over;cwer-
'overtemperature trip and those transients relying on other trip functions.
Those transients tripped on overpower-overtemperature are:
1.
Rod bank withdrawal at pcwer, 2.
Loss of load /.urbine trip, 3.
Coo 1~down' caused by feedwater system malfunction, 4.
Excessive load increase, 5.
Depressurization of Reactor Coolant System.
Those transients relying on other trips are (this list differs from that given in the :::' cal rs :rt and should su:er:ede it):
6.
Less of fhw (;ar-ial and complete loss),
7.
Startue of an inactive loep, S.
Dropped or misaligr.ed r:o.
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For the transients tripped on overpower-overtemperature, ' he DNB protection
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t trip setpoints are established ba, sed on THINC-IV calculated core limits and 4
are independent of the transient code calculations.
The DNB related parameters, power (AT), TAVG, and pressure are continuously
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monitored in the plant by the protection system and ill cause a plant trip as these measured parameters approach the core limits. These transients are also often predicted to reach other reactor trips such as high or low pressurizer pressure or high neutron flux. Any uncertainties in the calcu.
la.ted result would result in the DNB related parameters traversing a slightly different path and tripping the reactor if the ecmbination of parameters
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approached the core limits.
In the ONB evaluation of loss of flow events, no credit is taken for either the ' increase in system pressure or the eventual decrease in inlet temperature which result for these transients; these parameters are held constant at initial values in THINC for the ONBR calculation. The flow transients are calculated with conservative values'of pump inertia and are directly verified by plant measurements on each plant.
These data are presented in start-up test s.ummaries for each plant.
Input values listed in Table 2 provide a real but unquantified degree of conservatism in the calculation of heat flux; in. addition, comparison of best estimate calculations with reactor trip testsU) shows that LOFTRAR beunds the heat flux transient for high power bundles.' Unless the method of analyzing this event is changed, no code uncertainties beyond those suggested later by the staff need be
- riidersd.
Startup of an inactive loep is initiated from N-1 loep operating conditicns, i
generally 70% power for a four leco plant.
Fr:tec:fon is provicec by an IB m
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power setpoint and the overtemperature AT trfp which will have, been adjusted l
to N-1 loop operation values.. Typically; no credit is taken for the 85" l
power or the AT trips. Trip is assumed to occur on high neutron flux at 118". power; A considegable margin to the DNBR limit exists even with
" th;se conservative assumptions.
The DNB calculation for dropped or misaligned rods is a static THINC
' calculation which takes.no credit for plant response predicted by the sys'tems' transient code. The misaligned rod situation is simply a static ih' DNBR evaluation for the dropped rod case is a THINC-IV calculation event.
e perfonned assuming the maximum powershoot but taking no credit for the reduced core average temperature which would simultaneously be required for the overshoot to occur. Startup tests confirm the capability for calculating' the power overshoot.
Because neither the code uncertainties nor the offsetting code conservatisms
^
have been adequately quantified by Westinghouse, the staff will require that an uncertainty of 4". in DNBR be implemented in the design procedure for the 1
l.THINC-IV code uncertainty; and that, similarly, an uncertainty of 1".
in DNBR be implemented in the design procedure for transient code uncertainties. These uncertainties,which can be treated as 2e values, are to be used in equation (3)
(equation 3-8 of the topical report]for the calculation of the coefficient of variation. Westinghouse has,.nrasented qualitatWe arguments that these code uncertainties are not required; quantitative evidence is needed for the removal of these uncertainties. There are sev.eral ;:essible sources of TH::!C-IV uncertainty which have not been quan-ifi,e: cy 'iestinghouse.
These include:
l 1.
Uncertainty due to uncertainty in t.vo-phase friction factor i
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I correlation (Note: maximum friction, a function of voids, sets location of niinimum CNBRf, 2.
Error in nonnalization of axial shape, 3.
Use of infet flow boundary condition; 4.
Crossflow resistance independent of void fraction, 5.
Age effects on grids, 6.
Effect of non-unifonn clad s'urface roughness (crud effects),
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7.
Use '5fI5%,5fuit.io.'n_ j,nTinlEtiffoF_Jn_,,pl,[ce if t3D1ok y,6 pressure]
drop calculation.
The staff believes that compari.s'en of best estimate calculations with appropriate data, such as plant transient data, and quantification of design and input conservatisms could demonstrate that the imposed code uncertainties are
.not required. Th e re fo re th e. 's'til f f.'wi l 1'_ remo ve 'th e s eIs'ta f f l m'cos'ed
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uncertainties if sufficient conservatism is demonstrated by 'destinghouse.
Equation (2) is an empirical correlation of code calculations.
It is given as equation 3-2 in the topical report and is the model upon which the method is based.
Its applicability is subject to verification over the entire range of application.- That equation (2) provides a reasonable model of code calculations over a wide range of conditions was indicated in reference (5).
However, the sensitivity factors are dependent upon the particular correlations
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used in the code and the method of applying the code. Any changes in DNB correlation, or other assumptions or correlations used in the code require re-evalsation of the sensitiv*ty fact 5rs and of the model (ecuation (2)).
In reference (6), 'destinghouse stated "The applicability of the sensitivity factors, variances and means used in the analysis will be determined on a
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l plant-by-plant basis in the SAR". Also, the model (equation 2) must be I
justified each time a 'correlatioh in the code is changed.
For example, the model and the sensitivity factors in the topical report were obtained with the W-3 "R-grid" correlation; use of the WRB-1 correlation will require a new set of sensitivity factors and re-evaluation of the basic model. The effects of changes in DNB correlations or of correlations in THINC-IV should be
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submitted to_ permit the staff 'to r'e-e' valuate the basic me.de.l.
Sensitivity...-
factors used for a particular plant should be, included in the Safety Analysis Report or reload submittal.
Because equation (2) is an empirical fit to code calculations, rather than to measured data, errors in the correlation are fixed rather than random.
Therefore, the uncertainty allowance for equation (2), if required, should be. treated as an additive penalty in DNBR.
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i Asacheckonequation(2),Westinghouseanalyzedajairoftestcases I
i at their best estimate or nomi$al conditiolns and the'oth with the THINC-IV code.
values at their design or extreme conditiolns.
1 The ratio of the minimum DNBRs for the two test cases was 0.50 as determined by the THINC code.
'The minimum DNBR for the second case was,also calculated from the value
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for the first case by the relationship:
b XS L
. ( N') e( b)2-. +ko'hn3 v
i (6)
I where the p[ and ( represent the pai meter values from the first and second cases, respectively.
[2quation(6)canbeobtainedbyintegrating equation (2) and assuming 'that the sensitivity factors, S, are constants; g
the' ratio of the results as applied to two cases yields equation (6).
A value of 0.48 for the ratio of minimum (IN5Rs was predicted with equation (6).
I This indicates that, with the sensitivit'y factors u:;ed; equation (2) provides a conservative model for changes in DNSE, as caiculated by THINC, for small I
changes in parameter values. Therefore.,no uncertainty allowance is required for equation (2) with the sensitivity parameters used in the topical l
l report..Hewever, if the sensitivity factors are change 3 as a result of correlation changes, et, cetera, the uncertainty allcwance must be re-eval uated.
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Equation (3) is obtained by neglecting terms of second order and higher in a Taylor's series expansion of the integrated form of e:.a i:n'!). yestingh use :ss:ed ::e vCty Of this assum;;i:n for a limited range of parameter values.
The value of eacn para-meter was chosen :: be ene standarf deviaticn fecm its nominal is.
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l was ussd to determine the combined effect on DNBR. The calculation resulted in a ONBR value which was 0.79. times the best estimate value.
A Taylor's series with second and higher order terms neglected yielded a ONBR value which was 0.76 times the best estimate value.
The range tested for each parameter (1.7") is' insufficient to test the linearity over the range required but it does show that the linearity assumption combined with conservative values of the sensitivity factors y'ields conservative results for the limited values tested. The linearity should be tested over a wider range (12(T~as a minimum) each time the sensitivity factnrs are changed due to correlation changes, et. cetera.
Also, the comparison should be made at several values of 4
the parameters, e.g.,1T/2,17", i 2 T, to vecify linearity. Alternately;
' points might be chosen according to a factorial design or fractional replicate.
For the case presented in the topical report, a Monte-Carlo procedure which involved a random sampling of 100,000 cases from the variable distributions in-dicated a normal distribut' ion with mean and standard deviation that agree.within 1% of the values obtained using equation (3).(0} Comparisen of the Monte-Carlo procedure with results from equation (3) also provides an acceptable means of' testing linearity for future changes in sensitivity factors.
The proposed desig,n basis is that there must be at least a 9B% probability that the minimum DNBR. of the limiting power. rod is greater than or equal to the DNBR limit o'f the cor' relation being used; parameter uncertainties or variances obtained from the evaluation of data are determined at a 95P. con-ficer.ca level.
Im:lemen:a:icn Of :ne :r::*fure invcives assur::iers :na: er : -
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15-in correlations or postulated functions are random variables whereas repeated I
use of a correlation with identica,1 conditions gives identical results. Al so,
I distributions of uncertaintics for variables such as power, flow or temperature are not well known iso that the functional forms of the uncertainty distributions i
must be assumed.
Therefore, a rigorous statistical statement of the type implied in the proposed DNBR design basis cannot be obtained.
However, Westinghouse has either chosen distribution functions which are typical of observed distributions or which give conservative variances.
Also, biases in correlations have been reduced insofar as practical. Therefore, although a rigorous statistical :tatement cannot be made, the Westinghouse method provides a reasonable. approximation to such a statistical statement.
Variances and distributions for input parameters must be justified on a plant-by-plant basis until a generic approval is obtained.
In particular, the variance for primary coolant flow rate presented at a'95 percent upper confidence level in the topical report is incorrect because it was based on the wrong sample size. Also, the design-flow used in transient analyses must not exceed measured flow.
Parameters such as core inlet ficw maldistribution or outlet pressure distribution will also be dependent on the specific plant configuration. Distributions for parameters such as core average temperature will depend on the method of plant operation.
The discussion presented for rod bow is obsolete." DNB ' test data 'for 'ods '
r bowed to contact in test bundles containing simulated guide tubes show the rod bew effect to be larger than de:E nine:
f x crevicus tests.
An acceptable method of treating rod bow is 'to cmit it frcm the statistical treatment and add a rod bcw penalty directly to the :'NBR limit.
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Transients which.may be limited by parameters other than DNB must be analyzed with initial conditions appropriate.to those transients. For t
example, transients which approach code pressure limits should not be analyzed with nominal initial conditions for the peak' pressure evaluation.
In classifying tranisents as Condition I-IV events, Westinghouse considers the complete loss of coolant flow event to be,a Condition III event although they show it to satisfy Condition II event and requires that it satisfy Condition II criteria as presented in the topical report.
Table 3 lists all events which must satisfy Condition I or II crit' ria; e
any of these events which are not analyzed by the new proposed design procedure must satisfy the DNBR criteria of the more conservative fixed value method.
Staff Position The procedure for cal.culating DNB limits, as, presented in WCAP-8567, is acceptable for use in licensing applications.
It provides a reasonable approximation to the proposed statistical basis.
Certain restrictions must be imposed on the implementation of the method because of the sensitivity of the method to changes.in correlations used in design and because some of the details of the analysiJ in the topical report i
l are unacceptable.
These res'trictions are:
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Sensitivity factors used for a particular plant and their' ranges of applicability should be included in the Safety Analysis Report e
or reload submittal.
ThIsensitivityfactorsgiveninWCAP-8567 are acceptable provided the W-3 correlation is used and the parameter ranges of Table 1 are not exceeded.
2.
Any changes in DNB correlation, THINC-IV correlations, or parameter values listed in Table 1 outside previously demonstrated acceptable ranges require re-evaluation of the sensitivity factors and of the use of equation 3-2 of the, topical. report.
3.
If the sensitivity factors are changed as a result of correlation changes or changes in the application or use of the THINC code, then the use of an uncertainty allowance for application of equation 3-2 must be reevaluated and the linearity assumption of equation 3-3 of the re' port must be validated.
4.
Variances and distributions for input parameters must be justifie'd on a plant-by-plant basis until generic approval is obtained.
5.
Rod bcw cannot be treated as described in the topical report.
An acceptable approach is to add a rod bow penalty to th'e DNBR limit.
6.
Nominal initial condition assumptions apply only to DNBR analyses t
l using the new method. Other analysis, such as overpressure calculations, 1
require the appropriate conservative. initial condition assumptiens.
7.
Nominal conditions chosen for use in,, analysis should bound all permitted
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methods of plant operation.
8.
Code uncertainties, specified in th_ staff evaluation (+
4". for THINC-I'! and,15 for transients) must be ir,ciuded in the SBR ar.aly!'-
with the new method.
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Within the bounds of these constraints, the Westinghouse " Improved Thermal Design Procedure" is acceptable for use in licensing applications.
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The statistical method as presented includes no explicit design margin to j
accommodate unknoJns.
Such a margin could reduce or eliminate the impact of core related problems which are discovered after a core is designed and after a plant is operating. Although, no particular margin is quantified, 3
margin is inherent in the overall procedure used with the " Improved Thermal Design Procedure"; this margin is available to offset the effects of yet-to-be-discovered design problems. However, if newer procedures are proposed which ~substantially reduce the thermal margin, then a design margin to accommodate unknowns should be explicitly identified.
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The parameter ranges in Table 1, which are the parameter ranges for the sensitivity factors, do not cover the range required for part loop operation.
If the method is to be used for analysis of part loop operation, then the topical report must be amended to cover this wider range.
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References i.
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l 1.)
L.E.,Hochreiter and H. Chelemer, " Application of the THINC-IV Program j
to PWR Design," WCAP-8195, September,1973.
i 2.) letter NS-CE-1473 and Attachments, C. Eicheidinger to John F. Stolz, June 29, 1977.
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3.) letter NS-CE-1583 and Attachments,. C. Eicheldinger to' John F. Stolz, October 25, 1977.
4.)
T.M. Burke, C.E. Meyer and J. Shefcheck, " Analysis of Data from the Zion (Unit 1) THINC Verification Test", WCAP-8453-A, May,1976.
5.)
Letter NS-CE-1298 and Attachment, C. Eiche1dinger to John F. Stol'z,-
December 1,1976.
6.. )
Letter NS-CE-1060 and Attachment, C. Eiche1dinger to D.B. Vassallo, May 6,19 76.
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g Table 1 Demonstrated Applicable Parameter Ranges for Sensitivity Factors
..9 Parameter Ryg ll Reactor Coolant Flow 60% to 107% of nominal flow
'l (Assembly) 1.0 to 1.61 H
N F
(Rod) 9.0 to 1.61 AH Coolant Inlet Temperature 540*F to 640*F System Pressure
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1775 psia to 2400 psia Core Power 80% to 130% nominal l
Effective flow fraction same as reactor coolant flow E
-F 1.0 to value for fixed value method aH, 1 4
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Table 2 i
Typical Conservative Input Parameters Used in Transient Codes With Improved Thermal Design Procedure !
j Parameter SAR Value Expected Value i
Instrumentation
'I Response Time OTAT trip 6 sec 5.0 see -
High Neutron Flux 0.5 sec 0.3 see High Pressurizer Pressure 2.0 see 1.0 see Low Pressurizer 7.mssure 2.0 see 1.0 see Low Flow 1.0 see -
0.9 see Undervoltage 1.2 sec
.1.1 see Rod Drop Time 2.2 see 1.5 sec (toDashpot)
Moderator Density Coeff.
Limiting High Value 0.43ak/k/gm/cc 0.21 limiting Low Value 0 ak/k/gm/cc 0.068
. Doppler Power Coeff.
Limiting High Value 1.6%
1.2%
Limiting Lcw Value
.84%
1.0%
l Fuel Initial Tamperature Proprietary Reactivity Insertion Curve following trip Based'on Power Dist.
Bottom Approximat-ly Peaked Power Symmetric Dist.
Trip Reactivity Worth 4%
6%
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Table 3
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Condition I and II Events s
j 1.
Normal Operational Transients i
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2.
Uncontrolled RCCA bank withdrawal from a suberitical condition 3.
Uncontrolled RCCA bank withdrawal at power 4.
R,CCA misalignment 5.
Uncontrolled boron dilution o
6.
Partial or complete loss of forced reactor coolant flow 7.
Startup of an inactive reactor coolant loop 8.
Loss of normal feedwater 9.
Loss of external electrical' load and/or turbine trip 10.
Excessive heat removal due to feedwater system malfunction 11.
Excessive load increase in ident 12.
Loss of offsite power to station auxiliaries 13.
Steam pressure regulator malfunction 14.
Inadvertent opening of a steam generat'or relief or safety v lve
(
15.
Inadvertentclosure'ofmainsteamisolatiInvalves 16.
Inadvertent loading and operacion of a fuel assembly in an improper position i
i l
~
7
(
l Table' 3 (Cont'd) 18.
Inadvertent operation of ECCS during power operation A
l 19.
Chemical and Volume control system malfunction 20.
Inadvertent opehing of a pressurizer safety or relief valve O
S 6
I e
9 e
9 9
l e
(
G 6
i -
s
.9 90 i
1 7
~
6 5
v>
=o aa 33
= -i.it, a
u w
i = 2. 7 T.
m M
C wa 3
M w
e
~
=_:
=
2 t
- l...
.. -l.......-. L.
1 t
.1
.. _ L.. _ _1...
0 7
-6
-4
-2 0
2 4
6 l
l PERCEN Or FF ERENCE IN PRE')lCil0 AND MEA $i) RED FLOW (ICCr(P-M)/H) c o
i 1
1 i
71 j
'.te m.rtJ 7.edic s. I',: ar t F:c.v O:mce*,sen fer
- 3. s '.'. 4t : y.::; c ':. : s y h.se %nt s m-m e.
,,,,g g
7
^ j F
, ' CURE 2 DflBR LIMIT VALUE CALCULATION ii
(/t-1.645 f-)
1.30/F, cell type
@/t N
[
F, Limit DNBR n
typical
.0912 0.992
.0904 0.843 1.54 thin 61e
.0642 0.995
.0639 0.890 1.46
- 1ess than 1.0 because a " half-normal" distribution function is used for rod bow i-e i
s
\\
DNBR variation for thimble cell considering..
y variable values of. design parameters l
f DflBR (variable) = 1.30 Limit OilIR = 1.46 i
l l
l
-