ML20135A930

Nonproprietary Suppl 1 to Statistical Combination of Uncertainties, Parts 1-3

ML20135A930
Person / Time
Site:	Fort Calhoun
Issue date:	08/31/1985
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML19269B662	List: ML20135A898 ML20135A930
References
CEN-257()-NP, CEN-257()-NP-S1, CEN-257(0)-NP, CEN-257(0)-NP-S01, NUDOCS 8509100317
Download: ML20135A930 (42)

Text

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Cal-257(0)-NP SUPPLEMENT 1 -NP

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STATISTICAL  ;

COMBINATION .

OF .

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UNCERTAINTIES

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PART 1 AUGUST, 1985 R-

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E l POWER Biiiil SYSTEMS  :

l CCMBUSTION ENGINEERING. INC.

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LEGAL NOTICE This report was prepared as an account of work sponsored by Combustion Engineering, Inc. Neither Combustion Engineering nor any person acting on its behalf:

, A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefullness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.

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ABSTRACT

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This supplement updates Part 1 of the original report (CEN-257(0)-P, November, 1983) to account for the changes made to the Fort Calhoun Station Unit 1 TM/LP trip. Because this trip has been modified to include explicit monitoring of the shape index, the descriptions of the protection system and the analysis of the net uncertainties and the reported value of the net uncertainty have been changed.

The numbered sections of this report supersede the same numbered sections of the original report. Only Sections 2.3, 3.1, 3.2, Figure 2-1, and Tables 3-1, 3-2 and B1-lof the original report are changed.

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Table of Contents Section- Page Legal Notice i Abstract 11 Table of Contents iii Title Page 1 Report Sumary 2

' Supplement to Original Report 3 SECTION 2.3 3 SECTION 3.1 4 SECTION 3.2 4 TABLE 3-1 6 TARLE 3-2 7 TABLE B1-1 8 iii d i

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1 SUPPLEMENT 1 TO STATISTICAL COMBINATION OF UNCERTAINTIES PART I (CEN-257(0)-P) 9

REPORT

SUMMARY

Analysis of the DNB LSSS uncertainty factor accounting for the addition of the l ASI processing to the TM/LP trip follows almost the same procedure as used prior to the modification. That procedure was described in CEN-257(0)-P, Part

1. In this new procedure, a change has been made to the process for evaluating the allowance for TM/LP processing uncertainty. The addition of ASI monitoring to the TM/LP trip system allows [

3 and results in a new processing allowance of ( l (2 standard deviations) rather than the I _. .I allowance without the direct ASI processing. The net effect of these changes is that the uncertainty factor applied to the revised TM/LP system is f' 3 9

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2.3 TM/LP LSSS STOCHASTIC SIMULATION j l

For the TM/LP LSSS, DNB overpower (Pfdn) is the dependent variable of interest. Core coolant inlet temperature, reactor coolant system pressure, peripheral axial shape index, and total core power are monitored directly by the TM/LP trip system. Total integrated radial peaking factor and RCS coolant flow rate are monitored by other systems and must be included in the TM/LP LSSS

~

evaluations.

Figure 2-1 is a flow chart representing the simulation sequence for the TM/LP LSSS. For each simulation trial, a value of overpower obtained using sampled values of uncertainties about nominal conditions is calculated. This value is compared to the overpower calculated at nominal conditions by taking the ratio of the two values. This simulation sequence is repeated over several thousand sets of nominal operating conditions covering the operations space for the plant. The resulting distribution of the ratio of nominal overpower to overpower incorporating uncertainties is used to determine the overall uncertainty factor on the TM/LP LSSS.

i The operating space for the analysis is defined by establishing ranges for RCS pressure, fl ow, radial peaking (FR T), axial shape index (ASI) and core T

coolant inlet temperature. The ranges for FR and ASI are set to range of flow rate is set to cover flows from a minimum to a maximum value which might l occur with the present plant configuration. The pressure range is bounded by 4 the value of the high pressurizer pressure trip setpoint and the lower pressure limit of the TMfLP trip system. Core coolant inlet temperature covers a range established by the combination of core power and inlet temperature resulting in the highest temperature at which the secondary safety valves opean and the

lowest temperature at which the low secondary pressure trip occurs.

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The ranges of conditions used in the analysis are listed in Table 2-1.

1 Note: A line is drawn in the right hand margin where a change has been made.

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l 3.1 RESULTS OF ANALYSES The analytical methods presented in Section 2 have been used to show that a stochastic simulation of uncertainties associated with the Ap0 LSSS and the TM/LP LSSS results in combined uncertainties of f 1% and [ ]%, l respectively, at a 95/95 probability / confidence limit.

Table 3-1 shows the values of the individual uncertainties which were statistically combined to yield the above combined uncertainties. Appendix B contains a further discussion of the bases for these individual uncertainties.

The combined uncertainties are in units of percent overpower (P fdl and Pfdn) and are applied in the generation of the APD and TM/LP LSSS as

discussed below (Reference 3-1).

3.2- IMPACT ON MARGIN TO SAFDL The motivation for using a statistical combination of uncertainties is to improve NSSS performance through a reduction in the analytical conservatism in the margin to the SAFDL. This section contains a discussion of the margin obtainable through a reduction in this conservatism.

l Table 3-2 lists the uncertainty values previously used on this plant. The approximate worth of each of these uncertainties in terms of percent overpower margin (Pfd1' Efdn) is also shown.

The total uncertainties previously applied to the APD LSSS and the TM/LP LSSS are approximately ( .] and ( 'l , respectively. The uncertainties resulting from the application of the statistical combination of uncertainties program are approximately i 1 and ( ). The use of the statistical l combination of uncertainties provides a reduction in conservatism in the margin to SAFDL of approximately ( .I and ( l,respectively. l Although the conservatism in the margin to SAFDL has been reduced, a high degree of assurance remains that the SAFDL will not be violated.

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TABLE 3-1 UNCERTAINTIES ASSOCIATED WITH THE APD LSSS AND THE TM/LP LSSS Uncertainty

• APD LSSS ONR LSSS Core power (% of rated power) 121 12%

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Primary coolant mass flow (% design)** NA Primary coolant pressure (psid) NA Core coolant inlet temperature (FO ) NA l Power distribution (peaking factor) .

Separability See Table 1 of Appendix B1 Calibration (asiu)

Shape Annealing (asiu)

Monitoring system processing (asiu) i Monitoring system processing (psia) ,

Notes:

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• For complete description of these uncertainties, see Appendix 8.

• • Design flow = 190,000 GPM

• • • 2 o values

• • • • Includes f G 6

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TABLE 3-2 IMPACT OF STATISTICAL COMBINATION OF UNCERTAINTIES ON MARGIN TO SAFDL Approximate Values of Equivalent Overpowr Margin (%)

DNB APD Uncertainty Value LSSS LSSS Power 2% of rated Core coolant inlet 0

temperature 2F Reactor coolant system pressure 22 psid Axial shape index:

Separability .02 asiu Shape annealing .01 asiu Calibration .01 asiu Reactor coolant system flow 5460 gpm Peaking factors 5% DNB, 7% APD '

Equipment processing:

DNB LSSS 42 psid APD LSSS .02 asiu Total uncertainty applied Total previously Total uncertainty statistically combined

. Net margin gain , ,

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.. . - . . . _ . - . . . - - - . . .-. -.- - ---. ...- - . L:

TABLE B1-1 UNCERTAINTY [AND BIAS] COMPONENTS FOR THE EVALUATION OF THE PERIPHERAL SHAPE INDEX III K 95/95

, (asiu) K(f) (2) Bias I. Separability Uncertainty II. Calibration Uncertainty IRI III. Shape Annealing UncertaintyI "I IV. Monitoring System Processing

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Uncertainty LHR (asiu)

DN8R(psia)

, Notes on Table 1 (1) All components of the peripheral shape index have been tested for normality, and where indicated, satisfy that distributional requirement (n).

(2) f = degrees of freedom.

(3) This uncertainty is conservatively [ ,.

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(4) This'Ko95/95 is for consistent sets nf input data used by the uncertainty processors. .'

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CEft-257(0)-P, -NP, Suonlement 1 STATISTICAL COMBINATION OF UNCERTAINTIES .

PART 2 AUGUST,1985 e

E POWERSYSTEMS COMBUSTION ENGINEERING. INC.

L. - . .

LEGAL NOTICE . .

. This report was prepared as an account of work sponsored by Combustion ,

Engineering, Inc. Neither Combustion Engineering nor any person acting on f

. its behalf: ,

A. Makes any warranty or representation, express or implied including the  ;

warranties of fitness for a particular purpose or merchantability,  !

with respect to the accuracy, completeness, or usefullness of the >

- information contained in this report, or that the use of any l I information, apparatus, method, or process disclosed in this report

may not infringe privately owned rights; or t

l B. Assumes any liabilities with respect to the use of, or for damages

} resulting from the use of, any information, apparatus, method or process disclosed in this report. -

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i ABSTRACT

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. This supplement contains the revised MONBR limit for Fort Calhoun Nuclear Unit  !

1. This revision has been made based on NRC acceptance of the reduced (1.15)  !

CE-1 CHF correlation DN8R limit for C-E's 14x14 fuel. This supplement l demonstrates that there is at least 95% probability with at least 95%  !

i confidence that the limiting fuel pin will avoid departure from nucleate  !

boiling (DNB) so long as the MONBR found with the best estimate design CETOP-D >

model remains at or above 1.18. The 1.18 MONBR limit supersedes the value of 1.22 in the report CEN-257(0)-P.

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Table of Contents .

. Section Page Legal Notice i ,

. Abstract 11 Table of Contents 111 Title Page 1 Report Sumary 2 Supplement Text 3 1

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l Supplement 1 STATISTICAL CCMBINATION OF UNCERTAINTIES COMBINATION OF SYSTEM PARAMETER UNCERTAINTIES IN THERMAL MARGIN ANALYSIS FOR FORT CALHOUN NUCLEAR UNIT 1:

REVISED CHF CORRELATION UNCERTAINTY 4

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Report Sumary As described in Reference 2, the NRC has approved the CE-1 CHF Correlation, Part 2, which contains non-uniform axial power distribution CHF data. As a result, the itDNBR limit for the C-E 14x14 fuel with standard spacer grid is 1.15, higher than the 1.13 limit C-E had originally proposed.

The statistically derived 1.22 MDNBR limit reported for Fort Calhoun Nuclear Unit 1 in Reference 1 had included an NRC imposed interim penalty which increased the CHF correlation DNBR limit from 1.13 to 1.19, pending NRC's review of Part 2 of the CE-1 CHF correlation. Taking into account approval of a 1.15 CHF correlation DNBR limit, the new limit for Fort Calhoun Unit I has been determined to be 1.18, replacing the 1.22 limit previously reported in Reference 1.

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MDNBR Limit Pending Aporoval of CE-1 Correlation, Part 2 The MDNBR limit for Fort Calhoun Nuclear Unit I was 1.22 as reported in Reference 1. This limit included a penalty on the CE-1 CHF correlation, 1 pending NRC's review of the non-uniform axial power distribution data docu-mented in Reference 2. This penalty was included in the 1.22 limit as a shift of the mean of the DNBR p.d.f by 0.06, a conservative equivalent to a 5%

penalty on the 1.13 DNBR limit.

Adjusted MONBR Limit After completing a review of CE's non-uniform axial power distribution CHF data (Reference 2), the NRC concluded that a 1.15 DNBR limit should be used for CE's 14x14 fuel design. Since the 1.22 limit was generated prior to the l NRC's approval of the 1.15 CE-1 CHF correlation DNBR limit, it included an uncertainty penalty based on the NRC's interim CHF correlation DNBR limit of 1.19. The 1.22 limit can thus be reduced.

! The statistical combination of uncertainties analysis for Fort Calhoun Nuclear i Unit I resulted in a ONBR limit of 1.137 (p. 6-2 of Reference 1), which was I

increased to include a 0.5% allowance for fuel rod bow, a 0.06 (=1.19 - 1.13) i' CHF correlation uncertainty penalty and a TORC code uncertainty penalty of 1.1% as follows:

! 1.137 x 1.005 = 1.1427 '

1.1427 + 0.06 = 1.2027

. 1.2027 x 1.011 = 1.216 which was rounded up to 1.22, the final MONBR limit reported in Reference 1.

! The CHF correlation uncertainty penalty can now be reduced from 0.06 to 0.02 i

(=1.15-1.13). The revised DNBR limit is:

l

, 1.137 x 1.005 = 1.1427 j 1.1427 + 0.02 = 1.1627 1.1627 x 1.011 = 1.1755 which can be rounded up to 1.18.

CONCLUSION '

i Use of a 1.18 MONBR limit with a best-estimate design CETOP-D model for Fort Calhoun Nuclear Unit I will ensure with at least 95% probability and 95%

I confidence that the hot pin will not experience a departure from nucleate boiling. The 1.18 MONBR limit includes explicit allowances for system parame-ter uncertainties, CHF correlation uncertainty, rod bow, the NRC penalty for

, TORC code uncertainty and the revised penalty on the CE-1 CHF correlation.

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' REFERENCES i

! 1. ." Statistical Combination of, Uncertainties, Part 2", CEN-257(0)-P,

November, 1983.

j 2. "C-E Critical He,it Flux, Part 2: Non-Uniform Axial Power Distribution",

CENPD-207-P-A, December, 1984.

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CB1-257(0)-NP SUPPLEftENT 1 -NP 1

.I STATISTICAL. ' ~

COMBINATION i OF .

UNCERTAINTIES (

PART 3

\

l AUGUST, 1985 l '- .

i ed POWER i 9 SYSTEMS CCMBUSTICN ENGINEFING. INC.

l .

LEGAL NOTICE This report was prepared as an account of work sponsored by Combustion Engineering, Inc. Neither Combustion Engineering nor any person acting on e its behalf:

, A. Makes any warranty or representation. express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefullness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.

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ABSTRACT This Supplement updates the original report (CEN 257(0)-P, Part 3, November, 1983) to account for the addition of BASSS/ MINI CECOR to the DNB LCO monitoring system. Because this new system is based on continual monitoring of the power

.- distribution using in-core detectors, the description of the monitoring system, the analysis of the net uncertainties and the net uncertainty reported have e been changed.

The numbered sections of this report supersede the same numbered sections of the original report. In some sections significant additional material has been

- added. Sections 1.2.1, 1.4, 1.5, 2.3.2, 2.4.1.2, 2.4.1.3, 2.5, 3.1, 3.1.1,

3.1.2, 3.2.1, A.1, A.3, A.4 and Table 3-1 were changed. Figure 2-2 is added.

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Table of Contents Section Page Legal Notice i l Abstract 11

,; Table of Contents 111 Title Page 1 Report Sumary 2 Supplement to Original Report Section 1.~2.1 3 Section 1.4 4 Section 1.5 5 Section 2.4.1.2 8 Section 2.4.1.3 8 Section 2.5 9 Section 3.1 11 Section 3.1.1 11 Section 3.1.2 13 Section 3.2.1 13 Section A.1 16 Section A.3 17 Section A.4 18 Figure 2-2 10 Table 3-1 15 111 g . - - , . - _ . . - - - . - . - -

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SUPPLEMENT 1 TO STATISTICAL COMRINATION OF UNCERTAINTIES PART 3 (CEN-257(0)-P) ee 1

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REPORT

SUMMARY

Analysis of the DNB - LCO uncertainty factor with BASSS/ mini-CECOR follows the same procedure as the DNS - LCO uncertainty analysis described in CEN-257(0)-P, Part 3. Two changes in input uncertainties related to power distribution monitor'ng have been made. The [ ] processing uncertainty is no

,.' longer required and the only measurement uncertainty component for [. .

.] uncertainty. This uncertainty is given in CEN-

. 257(0)-P, Part - 3, App. B1, Table B1-1. All of the other non-ASI uncertainties remain the same. The total uncertainty factor is [ ).

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1.2.1 PROTECTION AND MONITORING SYSTEM The basic purposes and interactions of the LSSS and LCO's were previously described in Section 1.2.1 of Part 1 of this report. Part 1. describes the function of the protection system; Part 3 describes the function of the DNB and LHR LCO's.

Operation within the DNB and LHR LCO's provides the necessary initial DNB and LHR margin to prevent exceeding acceptable limits during Design Basis Events (DBE's) where changes in DNBR and linear heat rate are important. A list of the Nuclear Steam Supply System (NSSS) parameters which affect the calculation of these LCO's is shown in Table 1-2. A discussion of C-E setpoint methodology may be found in Reference 1-3.

Either the ex-core or the in-core detectors can be used to monitor the LHR ' C0 for C-E designed reactors. For Fort Calhoun Station Unit 1, the DNB LCO and axial :hape index can now be monitored on in-core detectors as well as on the ex-core detectors. This use of in-core detectors is described in reference 1-

5. Although these two DNB LCO monitoring systems are functionally similar in that they correlate allowed power levels and axial shape indices, use of the in-core detectors rather than the ex-core detectors implies different [*

__ ]. These differences are noted herein.

Note: A vertical line is drawn in the right hand margin where a change has been made..

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1.4

SUMMARY

OF RESULTS The analytical methods presented in Section 2.0 are used to show that a stochastic simulation of uncertainties associated with the ex-core detector-monitored DNB and LHR LCO's results in aggregate uncertainties of (* 'I and

( .1, respectively, at a 95/95 probability confidence level. The total

,. uncertainties previously applied to the ex-core DNB and LHR LCO's are approximately i 1 and [ J, respectively. Therefore, the statistical

.. combination of uncertainties program provides a reduction in the conservatism of the uncertainties applied in establishing the ex-core instrument monitored DNB and LHR LCO's of approximately ( l and [' 1, respectively. The stochastic simulation of uncertainties associated with in-core monitoring of the DNB LCO results in an aggregate uncertainty of ( l at the 95/95 probability / confidence level.

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1.5 REFERENCES

FOR SECTION 1 l 1-1 CEN-270-(0)-P, " Statistical Combination of Uncertainties," Part 1, November 1983, 1-2 CEN-270-(0)-P, " Statistical Combination of Uncertainties," Part 2, November

.! 1983.

/ 1-3 CENPD-199-P, Rev.1-P, Rev.1-P, "C-E Setpoint Methodology." March 1982.

14 Letter, E. G. Tourigny (NRC) to W. C. Jones (OPPD) dated March 5, 1983, License Amendment 70 and SER for Cycle 8 Operation of Fort Calhoun Station Unit No. 1, Docket No. 50-285.

1-5 CEN-119(B)-P, BASSS, November 1979. l l

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r 2.3.2 DNB LCO STOCHASTIC SIMULATION i

For the ONB LCO, ONB ove$ wer (Pfdn) divided by the required overpower margin (ROPM) is the dependent variable of interest. the core coolant inlet temoerature, reactor coolant system pressure and flow rate, peripheral axial shape index and integrated radial peaking factor are the independent variables

, of interest. As demonstraced in Appendix C, R0PM is relatively insensitive to-these independent variab'.es. In addition, the maximum R0PM -as a function of

j s shape index is used as input to generate the LCO's.- This reduces the analytical evaluation of the dependent variable to consideration of the Pfdn's response to the uncertainties of the independent variables. TORC /CE-1 (References 2-2, 2-3) is used to determine the functional relationship between p P fdn and the independent variables. The probability distribution of i uncertainties associated with some of the independent variables have been discussed in Appendix A of Part 1 of this report. Those uncertainties

, specifically associated with the calculation of the core average axial shape index us'ing the in-core detector system to monitor the DNB LCO are discussed in l Appendix A of this part of the report.

The core coolant inlet temperature range of interest for the DNB LCO stochastic simulation is defined by:-

j (1) the temperature at which the secondary safety valves open, and l (2) the temperature at which the low secondary pressure trip would occur The reactor coolant system pressure range of interest for the DNB LCO stochastic simulation is defined by:

(1) the value of the high pressurizer pressure trip setpoint, and i .-

(2) the lower pressure limit of the thermal margin / low pressure trip 4

, It is noted that these ranges are the same as used in the LSSS stochastic simulation (Ref. 2-1) and as such are bounding for the LCO.

Figure 2-1 is a flow chart representing the ex-core detector monitoring stochastic simulation of the ONB limits. This- figure is similar to Figure 2-1 in Part 1. Figure 2-2 is a flow chart representing the in-core detector

' monitoring stochastic simulation of the DNB limits.

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This figure differs from 6

Figure 2-l'in that the [

]

stochastic simulation. The independent variables and their uncertainties are input to CESCU. Each data set generated.by the statistical part of CESCU is evaluated with the CETOP-D portion of CESCU to generate a Pfdn probability distribution.- The ratio of the mean value of P fdn to the lower 95/95 value

,7 of P fdn is the parameter of interest. The details of the specific DNB LCO stochastic simulations performed are presented in Section 2.4.-

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2.4.1.2 AXIAL SHAPE INDEX UNCERTAINTY SIMULATION  ;

2.4.1.2.1 EX-CORE AXIAL SHAPE INDEX At Fort Calhoun Unit 1, the digital display of safety channel information is used tomonitor the LHR and DNB LCO's, in accordance with Technical I* Specification 2.10.4. Thus, the basic relationships between the components of the axial shape index uncertainty for LSSS, described in Appendix B1 of Part 1, are also appropriate for the LCO uncertainty analysis.

2.4.1.2.2 CORE AVERAGE AXIAL SHAPE INDEX The uncertainties associated with the in-core detector system have been developed in support of the better axial shape selection system (Reference 2-5).

The magnitude of those uncertainties are defined in Appendix A of this part of the report.

The procedure used to sample the shape index uncertainty distributions for the LCD stochastic simulation are those described in Section 2.4.1.2 of Part 1.

2.4.1.3 PROCESSING UNCERTAINTY SIMULATION 2.4.1.3.1 EX-CORE INSTRUMENT PROCESSING The signals generated by the ex-core detectors are processed into an axial shape index (ASI) value. The electronic processing equipment introduces further uncertainty in these values. However, the safety channel information used is transferred to the digital display before the information is processed by the trip actuation evaluation circuits. Therefore, only the safety channel processing uncertainties need by included in the LCO processing uncertainties.

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, Since the axial power distribution and the ASI value used in each simulation calculation are correlated, this uncertainty is incorporated in the stochastic

,. evaluation of the LCO.

2.4.1.3.2 IN-CORE INSTRUMENT PROCESSING As noted in Appendix A the processing uncertainty for the in-core instrument signals has been [

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2.5 REFERENCES

FOR SECTION 2 2-1 CEN-257(0)-P, " Statistical Combination of Uncertainties," Part 1, November 1983.

2-2 CENPD-161-P, " TORC Code: A Computer Code for Determining the Thermal

$ Margin of a Reactor Core," July,1975.

j 2-3 CENPD-206-P, " TORC Code: Verification and Simplified Modeling Methods,"

Janua ry,1977 2 A C. Chiu, J. F. Church, "Three-Dimensional Lumped Subchannel Model and Prediction-Correction Numerical Method for Thermal Margin Analysis of PWR Cores," TIS-6191 June, 1979.

2-5 CEN-119(B)-P, BASSS, November 1979.

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c 3.0 RESULTS AND CONCLUSIONS 3.1 RESULTS OF ANALYSES The statistical analytical methods presented in Section 2 have been used to show that a stochastic simulation of uncertainties associated 91th the ex-core l

monitored DNB and LHR LCO result in combined uncertainties of ( .I and

[ ], respectively, at a 95/95 probability confidence level. Stochastic

" simulation of the in-core monitored DNB LCO results in an aggregate uncertainty of ( ).

Table 3-1 shows the values of the individual uncertainties which were I statistically combined to yield the above combination. Appendix A contains a

further discussion of the bases for these individual uncertainties.

i The combined uncertainties are in units of percent overpower (P fdn, P fdl I and are applied as such in the generation of the LCO limits as discussed below' i

(Reference 3-1).

3.1.1 DNB LCO I The fuel design limit on DNBR for the DNB LCO is represented by a combination j of the ordered pairs (Pfdn, ASIONB). A lower bound is drawn under the

" flyspeck" data such that all the core power distributions analyzed are ,

- bounded. This lower bound is reduced by applicable uncertainties as follows:

I e i (3-1) i b '.

(3-2)

I (3-3)

• Equations 3-2 and 3-3 are valid for the excore and incore monitoring systems, respectively.

11 4

- ,, . _ _ _ _ - _ _ _ _ _ _ . . _ . . _ _ . , , _ _ , , - . _ . . - , , . __.,7 .

where:

LCO BOMB DNB Power limit for LCO after inclusion of uncertain-ties and allowances P

fdn

- Power to fuel design limit on DNB including the effects of azimuthal tilt

- SMD0 - Statistically combined uncertainties applicable to the DNB

, LCO ASI ONB - Axial shape index associated with Pfdn L

r Temperature, pressure and flow components of the DNB LCO are represented by '

equations as folows:

l (3 4)

! (3-5) l

,, . (3-6) l where:

FDNB. PDNB, TDNB = Coolant conditions used in the J

in calculations of (P fdn' I) p

ordered pairs of data.

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, 12

3.1.2 LHR LCO l The excore detector monitored LCO on linear heat rate is represented by the ordered pairs (P fd1' Ip ). A lower bound is drawn under the "flyspec" data such that all the core power distributions analyzed are bounded. This lower ,

, bound is reduced by the applicable uncertainties and allowances to generate the LCO as follows:

l 3 ._

(3-7) }

l l (3 8) J I

j where:

Bh - Linear Heat Rate Power Limit for LCO after inclusion uncertainties.

PM-ThepowertotheLCOlinearheatratelimitincluding i the effects of azimuthal tilt.

SMLO -

Statistically combined uncertainty applied to the LHR LCO The incere detector monitored LC0 on linear heat rate will not be modified for a statistical combination of uncertainties.

3.2 IMPACT OF STATISTICAL COMBINATION OF UNCERTAINTIES l

3.2.1 IMPACT ON MARGIN

.. The motivation for using a statistical combination of uncertainties is to improve NSSS performance through a reduction in analytical conservatism in the

uncertainties which must be taken into account. This section contains a discussion of the margin obtainable through a reduction in this conservatism.

1 13

i l

Table 3-2 lists the uncertainty values previously used on Fort Calhoun Unit 1.

! The approximate worth of each of these uncertainties in terms of percent j overpower margin (Pfdn' Efdl) is shown.

The total uncertainties previously applied to the excore monitored DNB and LHR LCO are approximately I .1 and I. 1, respectively. The use of the statistical combination of uncertainties justifies a reduction in the conservatism in the uncertainty of approximately I ] and ( 1,

.T respectively. The use of the statistical combination of uncertainties and incore detector monitoring of the DNB LCO results in an uncertainty of approximately I. 1.

Although the conservatism in the uncertainty has been reduced, a high degree of assurance remains that acceptable limits will not be exceeded.

95 9

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1 14

r TABLE 3-1 UNCERTAINTIES ASSOCIATED WITH THE DNR AND LHR LCO'S Uncertainty LHR LCO ONB LCO a

Core power (% of rated power) +2% +2%

$ Primary coolant mass flow (% Design flow)* NA ( )

Primary coolant pressure (psia) NA ( ')

Core coolant inlet temperature (FO ) NA [ ] l Power distribution (peaking factor) ( 1 Axial Shape Index (Excore Detector System) *l

1. Separability (asiu) See Table A-1 of Appendix Al

2. Calibration (asiu) (,, J ( )

! 3. Shape Annealing (asiu) ( )

4 Monitoring system processing ( .]

(asiu) (2a) ,

Axial Shape Index (Incore Detector System) i 1 ,

(asiu) l

, Note: For complete description of these uncertainties, see Appendix A.

• Design Flow: 190,000 gom f

15

l A.1 SHAPE INDEX UNCERTAINTIES A.1.1 AXIAL SHAPE INDEX UNCERTAINTIES ASSOCIATED WITH THE EXCORE DETECTOR SYSTEM At Fort Calhoun Unit 1, the safety channel instruments, which supply some of

, the input information to the trip system, also supply information for monitoring the shape index of the LCO's. This shape index information is i' supplied to the LC0 monitors before the trip system evaluation of shape index is performed. Consequently, except for the processing uncertainty for shape index, uncertainties described for the LSSS in Appendix A-1 Part 1 (Reference A-1) are appropriate for evaluation of the LCO shape index uncertainty. The LCO shape index processing uncertainty should include only that part of the LSSS processing uncertainty attributed to the ex-core detectors, and must exclude that portion of the uncertainty due to the trip system shape index evaluation circuits. The processing uncertainties are described in A.3 below.

The other shape index uncertainties are given in Table A-1.

A.I.2 AXIAL SHAPE INDEX UNCERTAINTIES ASSOCIATED WITH THE INCORE DETECTOR SYSTEM The incore detectors are used to calculate the core average axial shape index, I. The value of I calculated from the incore signals is used to monitor the DNB LCO in the BASSS/ Mini-CECOR system. Mini-CECOR is just a mini computer version of C-E's CECOR code. CECOR and its uncertainties are described in Reference A-2.

t

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16

A.3 MONITORING SYSTEM PROCESSING UNCERTAINTIES A.3.1 EXCORE MONITORING SYSTEM The description of the Trip System Processing Uncertainties given in Appendix B3

of Part 1 is valid for the Fort Calhoun excore monitored DNB t.C0 because the J

, same instruments are used.

I

$ A.3.2 IN-CORE MONITORING SYSTEM The processing uncertainties for the DNB 1,C0 in-core monitoring system result from the [ ,

s

3 1

2 1

a O

17

. . -_ -. .. .- _ . . ~ - - - . _ = .

i l

l l A.4 REFERENCES FOR APPEN0!X A l

A-1 CEN-2578-(0)-P, " Statistical Combination of Uncertainties Part 1,"

November, 1983.

f A-2 A. Jonsson, W. B. Torney, M. W. Crump " Evaluation of Uncertainty in the

, Nuclear Power Peaking Measuring the Self-Powered, Fixed In-Core Detector System " CEMPD-153-P, Rev. 1-7, May, 1980.

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