ML17228A271
| ML17228A271 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/27/1993 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17228A270 | List: |
| References | |
| NUDOCS 9309020213 | |
| Download: ML17228A271 (47) | |
Text
St. Lucie Unit 2 Docket No. 50-389 L-93-217 ST. LUCIE UNIT 2 MARKED-UP TECHNICAL SPECIFICATION BASES PAGES Page B 2-1 Page B 2-4 Page B 3/4 2-2 9309020213 930827'OR ADOCK 05000389 P
.P.OR....
2.1 SAFETY LIMITS BASES gpFPL +5 t,Qg ln cogjvAc+lor1 InlL+h
+he Exkended Qafiskica.( Gobi'ncdion of Qgc 4 i~+ es (t--SCU)-
- 2. 1. 1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting will occur.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (ONB) and the resultant sharp reduction in heat transfer coeffi-cient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to ONB through the CE-1 correlation.
The CE-1 ONB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the mar gin to ONB.
The minimum value of the DNBR during steady state operation, normal opera-tional transients, and anticipated transients is limited to This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-1 ONB correlation uncertainty.
This value corresponds to a
95K probability at a
95K confidence level that ONB will not occur and is chosen as an appropriate margin to ONB for all operating conditions.
The curves of Fi ure 2.1-1 show loci of points of THERMAL POWER, Reactor oolant System pressure and maximum cold leg temperature with four Reactor Cool-ant Pumps operating for which the mum=
o for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1.
The limits in Figure 2.1-1 were calculated for reactor coolant inlet temp'eratures less than or equal to 580'F.
The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature.
Reactor. operation at THERMAL POWER levels higher than 112% of RATED THERMAL POWER is prohibited by the high power level trip set-point specified fn Table 2.2-1.
The area of safe operation is below and to the left of these lines.
The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.
The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Variable Overpower Trip and the Power Dependent Insertion Limits, assure that the Specified Acceptable Fuel Design Limits on ONB and Fuel Centerline Melt are not exceeded during normal operation and design basis Anticipated Operational Occurrences.
ST.
LUCIE-UNIT 2 8 2-1 Amendment No. P,P D~e S~~o~
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Variable Power Level-Hi h A Reactor'.trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure Trip.
The Variable Power Level High trip setpoint is operator adjustable and can be set no higher than 9.61K above the indicated THERMAL POWER level.
Operator action is required to increase the trip setpoint as THERMAL POWER is increased.
The trip setpoint is automatically decreased as THERMAL POWER decreases.
The trip setpoint has a maximum value of 107.0X of RATED THERMAL.
POWER and a minimum setpoint of 15.0X of RATED THERMAL POWER.
Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 112K of RATED THERMAL POWER, which is the value used in the safety analyses.
Pressurizer Pressure-Hi h
Amendment Nn.f./O, The Pressurize~
Pressure-High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor t'rip.
This trip's setpoint is at less than or equal to 2375 psia which is below the nominal lift setting 2500 psia of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves.
+he Dgg"SQFDL ok l.2o> IH conjunc+ion Thermal Mar in/Low Pressure wi+4 $ 5CLL wekhodolctgy, The Thermal Margin/Low Pressure trip is provided to prevent operation when the ONBR is iess than~
The trip is initiated whenever the Reactor Coolant System pressure signal drops below either 1900 psia or a computed value as described below, swhichever is higher.
The computed value is a function of the higher of hT power or'eutron
- power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX.
The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.
In addition, CEA group sequencing in accordance with Specifica-tions 3. 1.3.5 and 3. 1.3.6 is assumed.
Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence p~io~ to a Power Level-High trip is assumed.
The Thermal Margin/Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error.
A safety margin is provided which includes an allowance of 2.0X of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 3.0 F to compensate for potential temperature measurement uncertainty; and a further allowance of 125 psia to compensate for pressure measurement error and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit.
The 125 psia allowance is made up of a 55 psia pressure measurement allowance and a 70 psia time delay allowance.
ST.
LUCIE.- UNIT 2 B.2-.4
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POWER 0ISTRIBUTION LIMITS BASES assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Oensity " High LCOs and LSSS setpoints remain valid.
An AZIMUTHAL POWER TILT > 0. 10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.
The requirement that the measured value of Tq be mutiplied by the calculated values of F and F
to determine F
and F
is applicable only xy xy when F
and F
are calculated with a non-full core power distribution analysis xy code.
When monitoring a reactor core power distribution, F
or F
with a full r
xy core power distribution analysis code the azimuthal tilt is explicitly accounted for as part of the radial power distribution used to calculate Fx
'and Fr.
The Surveillance Requirements for verifying -that F
F and T
are T
T xy' within their limits provide assurance that the actual values of F, F
and T
xy' do not exceed the assumed values.
Verifying F and F
after each fuel T
T xy r
loading prior to exceeding 75K of RATEO THERMAL POWER provides additional assurance that the core was properly loaded.
3/4.2.5 ONB PARAMETERS The limits on the ONB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of 'operation assumed in the transient and safety analyses.
The limits are consistent with the safety analyses assumptions nd ave been analytically demonstrated adequate to maintain a minimum ONBR of >
throughout each analyzed transient.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12-hour basis.
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LUCIE - UNIT 2 B 3/4 2-2 Amendment No.
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St. Lucie Unit 2 Docket No. 50-389 APPXMVIT K3RKIMT TO 10 CFR 2.790 Combustion Engineering, Inc: July 7, 1989 (4 pages)
- AFFIDAVITPURSUANT TO 10 CFR 2.790 Combustion Engineering, Inc.
)
State of Connecticut
)
County of Hartford
)
SS.:
I, A. E. Scherer, depose and say that I am the Director, Nuclear Licensing, of Combustion Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below.
I am submitting this affidavit in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations and in conjunction with the application of Florida Power and Light Company for withholding this information.
The information for which proprietary treatment is sought is contained in the following document:
CEN-371(F)-P, "Extended Statistical Combination of Uncertainties".
This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced
- document, should be withheld.
1.
The information sought to be withheld from public disclosure concerns the methodology,'pecific results, and an estimation of benefits of applying the extended statistical combination of uncertainties method for use in the
'etpoint calculations of DNB LSSS and LCO limits, which is owned and has been held in confidence by Combustion Engineering.
2.
The information consists of test data or other similar data concerning a
- process, method or component, the application of which results in substantial competitive advantage to Combustion Engineering.
3.
The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public.
Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in J
confidence.
The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stern to Frank Schroeder dated December 2, 1974.
This system was applied in determining that the subject document herein are proprietary.
4.
The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.
5.
The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
6.
Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:
I ~
~
3 a.
A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.
b.
Development of this information by C-E required thousands of manhours and hundreds of thousands of dollars.
To the best of my knowledge and belief a competitor would have to undergo similar expense in generating equivalent information.
c.
In order to acquire such information, a competitor would also require considerable time and inconvenience developing the method of extended statistical combination of uncertainties.
d.
The information required significant effort and expense to obtain the licensing approvals necessary for application of the information.
Avoidance of this expense would decrease a competitor's cost in applying the information and marketing the product to which the information is applicable.
e.
The information consists of the methodology used to statistically combine uncertainties, the application of which provides a competitive economic advantage.
The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.
f.
In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be included.
The ability of Combustion Engineering's competitors to utilize such
s
~
I information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.
g.
Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development.
In addition, disclosure would have an adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees.
Further the deponent sayeth not.
A. E.
erer Director Nuclear Licensing Sworn to before e
th.s a~day of Notary Public I
I
( My'ommission expires
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St. Lucie Unit 2 Docket No. 50-389 L-93-217 Combustion Engineering, Inc. Report K%ENDED STATISTICAL COMBINATION OF UNCERTAINTIES CEm-371(F) -X
- July, 1989 (Bound Document)
1
~
I PRRIETARY INFORMATION This Document contains proprietary infor.
mation and is not to be transmitted or re.
produced without specific written ap.
proval from Combustion Engineering, Inc.
Cepy No.
CEH-371(F) -P EXTENDED STATISTICAL COMBINATION OF UNCERTAINTIES JULY, 1989 c~~>+STION ENOINKKRINC PDR ADQCK 05000389 P
LEGALNOTlCK THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC.
NEITHER CQMBUSTlON ENGINEERING NOR ANYPERSON ACTING ON ITS BEHALF:
A.
MAKES ANY WARRANTY OR REPRESENTATION, ECPRESS QR IMPLIED INCI.UDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE
- ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATIONCONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, QR PROCESS DISCI.OSED IN THIS REPORT MAY NQT INFRINGE PRIVATELY OWNED RIGHTS; OR B. ASSUMES ANY LIABILITIESWITH RESPECT TO THE USE OF, QR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION,APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
CEN-371(F) -P EXTENDED STATISTICAL COMBINATION OF UNCERTAINTIES AN IMPROVED METHOD FOR STATISTICALLY COMBINING UNCERTAINTIES FOR THE C-E CALCULATED THERMAL MARGIN/LOW PRESSURE LSSS AND DNB LCO FOR ST.
LUCIE UNIT II.
ABSTRACT The Extended Statistical Combination of Uncertainties (ESCU) report describes an improved method for statistically combining uncertainties for the C-E calculated Thermal Hargin/Low Pressure (TH/LP)
LSSS and Departure from Nucleate Boiling (DNB) Limiting Conditions for Operation (LCO) for St. Lucie Unit II.
ESCU is a modification to the NRC approved Statistical Combination of Uncertainties (SCU) method currently in use for St. Lucie Unit II.
This report describes the ESCU method of calculating total uncertainties expressed in [percent overpower and its associated DNB SAFDL.
- Together, when applied in DNB LSSS and LCO setpoint calculations, they] assure with at least' 95% probability at a
95% confidence level that the hottest fuel rod in the core does not experience a Departure from Nucleate Boiling (DNB) during normal operation or an Anticipated Operational Occurrence (AOO) initiated within the LCO limits.
0 TABLE OF CONTENTS CHAPTER 1.0 Introduction PAGE 1.1 Purpose
1.2 Background
1.2.1 1.2.2 1.2.3 1-1 1-1
z I 0 W XI v) me CA CORE AVERAGE ASI ASI CORREC-TIONS Th Tc Ip ELECTRONIC PROCESSING UNCERTAINTIES hlp2 OVERPOWER vs I SENSITIVITY RELATION ASI hlP1 UNCERTAINTY PROCESSORS
~Bopm tIp)
MAXIMUM VALUEOF 5opm 595/95 OP Ill Pfdll~hBo m BMU POWER MEASUREMENT (BMU)
BUILDUP UNCERTAINTYDISTRIBUTIONON UNCERTAINTIES OVERPOWER FOR N CASES EO
STATE PARAMETER UNCERTAINTIES ASI UNCERTAINTIES ROCESSING UNCERTAINTIES STOCHASTIC SIMULATION APPROACH EQUIVALENTPENALTY FROM OVERPOWER p,d.f.
SETPOINT ANALYSIS SYSTEM PARAMETER UNCERTAINTIES CHF CORRELATION UNCERTAINTIES THERMAL HYDRAULIC
RESPONSE
SURFACE EQUIVALENTPENALTY FROM DNBR p.d,f.
FLORIDA POWER 4 LIGHT CO.
~St. Lucl~ 2 Nuclear Perm<
Plant SCU APPROACH Figure 1-3 1-7
STATE PARAMETER UNCERTAINTI ES ASI UNCERTAINTIES ROC ESSING UNCERTAINTIES STOCHASTIC SIMULATION MODEL EQUIVALENT PENALTY FROM OVERPOWER p.d,f, SETPOINT ANALYSIS SYSTEM PARAMETER UNCERTAINTIES CHF CORRELATION UNCERTAINTIES THERMAL HYDRAULIC
RESPONSE
SURFACE DNBR p.d.f.
FlORIDA POWER 5 L,IGHT CO.
St. Lucia 2 Nuclear Power Rant ESCU APPROACH Figure 1-4 1-8
Report Scope The objective of this report is to define the methods used to statistically combine uncertainties applicable to the Thermal Margin/Low Pressure (TM/LP) Limiting Safety System Settings (LSSS) and the Departure from Nucleate Boiling (DNB) Limiting Conditions for Operation (LCO).
The report encompasses the following issues:
1.
To define the new DNB LSSS and DNB LCO stoc(astic simulation models which calculate the overall uncertainty penalties.
2.
To define the
[DNBR probability density function] which is input to the stochastic simulation models.
3.
To evaluate the aggregate uncertainties and the
[associated DNBR SAFDL] as they are applied in the determination of the TM/LP LSSS and DNB LCO.
Summary of Results The analytical methods presented in section 2 are used to show that the application of the ESCU method with an [associated DNB SAFDL of 1.20] results in typical uncertainty penalties of
[7.5% overpower] for TM/LP LSSS and
[8.2% overpower] for ex-core DNB LCO.
References for Section 1:
l-l* "Statistical Combination of Uncertainties Part 1"
CEN-123(F)-P
- December, 1979..
1-2* "Statistical Combination of Uncertainties Part 2",
CEN-123(F)-P January, 1980.
1-3* "Statistical Combination of Uncertainties part 3",
CEN-123(F)-P February, 1980.
1-4 "CE Setpoint Methodology",
CENPD-199-P Rev.
1-P-A,
- January, 1986.
1-5
- Letter, C.
C. Nelson (NRC) to R.
E. Uhrig (FP&L),
"Ammendment ¹48 to Facility Operating License ¹ DPR-67 for St. Lucie 1, Docket ¹50-335,"
November 23, 1981.
These References have been approved for use by the NRC in Reference 1-5.
1-9
2.0 ANALYSIS 2.1 Objectives of Analysis The objectives of the statistical analysis are to determine overall uncertainty factors to be applied to the TM/LP LSSS and the DNB LCO.
These uncertainty factors are determined such that there is a
95% probability at a
95% confidence level that the combined effect of uncertainties on the LSSS and LCO limits will not exceed these factors.
2.2 2.3 Analytical Techniques The techniques used to evaluate, the uncertainty factors are similar to those used on CE plants employing Analog Reactor Protection Systems as described in References 2-1, 2-2 and 2-3.
The only functional change is to add the
[DNB SAFDL as a
distributed parameter, rather than a fixed value].
The CE SCU stochastic simulation methodology described in References 2-1 through 2-3 is used to determine the overall uncertainty factors.
TH/LP LSSS Stochastic Simulation For the TH/LP LSSS as described in Ref. 2-1, DNB overpower (Pfdn) is the dependent variable of interest.
Core coolant inlet temperature, reactor coolant system pressure, core power, and peripheral axial shape index 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 TH/LP LSSS evaluations.
[The value of the DNBR SAFDL is a system parameter determined from the DNBR p.d.f. discussed in appendix A and is included in the simulation as a
CETOP input].
Figure 2-1 is a flow chart representing the simulation sequence for the TH/LP LSSS.
For each simulation trail, a value of over power 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 ration of the two values.
This simulation sequence is repeated over a large number 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 TH/LP LSSS.
For the SAFDL, [the nominal value is set to the 95/95 upper limit of the CE-1 CHF correlation i.e.
1.20, rather than the mean value].
This is done to be consistent with the requirements of the Section 4.4 of the Standard Review Plan.
2-1
2.4 DNB LCO Stochastic Simulation For the DNB LCO, DNB overpower (Pfdn) divided by the required margin (ROPM) is the dependent variable of interest.
The core coolant inlet temperature, reactor coolant system pressure and flow rate, peripheral axial shape index and integrated radial peaking factor are the independent variables of interest.
The DNBR SAFDL is included as a distributed input in an identical fashion to the TM/LP LSSS.
Similar to the approach taken for SCU (Ref. 2-1), the maximum ROPM as a function of shape index is used as input to generate the DNB LCO.
This reduces the analytical evaluation of the dependent variable to consideration of Pfdn response to the uncertainties of the independent variables.
CETOP-D is used to determine the functional relationship between Pfdn and the independent variables.
2.5 The probability distributions of uncertainties associated with the independent variables have been discussed in Reference 2-3.
Figure 2-2 is a flow=chart representing the ex-core detector monitoring stochastic simulation of the DNB limits.
This figure is similar to Figure 2-1.
References for Section 2:
2-1* "Statistical Combination of Uncertainties, Part 1"
CEN-123(F)-P,
- December, 1979.
2-2* "Statistical Combination of Uncertainties, Part 2 January, 1980" CEN-123 (F) -P.
2-3* "Statistical Combination of Uncertainties, Part 3" CEN-123 (F)-P, February 1980.
2-4
- Letter, C.
C. Nelson (NRC) to R.
E. Uhrig (FP&L),
"Ammendment ¹48 to Facility Operating License ¹ DPR-67 for St. Lucie 1, Docket ¹50-335,"
November 23, 1981.
These References have been approved for use by the NRC in Reference 2-4.
2-2
PROBABILITY DISTRIBUTIONS N INPUT ARAMETERS SAMPLING MODULE Tc P
Fr F
FDL AXIALPOWER DISTRIBUTION P =P+BP C
E T
0 P
Pfdn REFERENCE
~ SAFDL zmZl XIA Z
CZ O
Z z
I CORE AVERAGE ASI ASI CORREC-TIONS Th Tc IP ELECTRONIC PROCESSING UNCERTAINTIES OVERPOWER vs I SENSITtVITY RELATION hl ASI P
UNCERTAINTY PROCESSORS
~Bopm (IP)
MAXIMUM VALUEOF 5opm 595/95 QP Al Pfdn+ABO m+BMU POWER MEASUREMENT (BMU)
BUILDUP UNCERTAINTYDISTRIBUTIONON UNCERTAINTIES OVERPOWER FOR N CASES Tl
~
(Qc tD
4 PROBABILITY DISTRIBUTIONS ON INPUT PARAMETERS SAMPLING MODULE Tc P
Fr F
FDL AXIALPOWER DISTRIBUTION C
E T0 P
Pfdn REFERENCE SAFDL zm U
g) X f-n nQ Q 3)m m
n+
mm
~n zz C Q
~z z
I
~Q
~m CORE AVERAGE ASI ASI CORREC-TIONS Th Tc Ip ELECTRONIC PROCESSING UNCERTAINTIES BIP2 OVERPOWER vs I SENSITIVITY RELATION ASI hlPl UNCERTAINTY PROCESSORS
~Bopm
. (Ip)
MAXIMUM VALUEOF Bopm B95/95 opm fdo+ABo m+BMU
~~
h) e POWER MEASUREMENT lBMW BUILDUPUNCERTAINTYDISTRIBUTION ON UNCERTAINTIES OVERPOWER FOR N CASES
3.0 RESULTS AND CONCLUSIONS 3.1 Results of Analysis Table 3-1 presents the TH/LP LSSS and DNB LCO statistically combined uncertainties from application of the ESCU analytical methods described in Section 2.
The individual uncertainties and their corresponding values which were combined are the same as presented in References 3-1 to 3-3 except for the addition of two uncertainty items to the HDNBR p.d.f.
(Described in Appendix A) which were added in the course of the NRC review.
3.2 The aggregate uncertainties are in units of [percent overpower]
and are applied as described in Section 3. 1.2 of Reference 3-1 for TH/LP LSSS and Section 3. 1. 1 of Reference 3-3 for DNB LCO.
Conclusion The Extended Statistical Combination of Uncertainties (ESCU) methods and typical results have been presented in this report.
This methodology will continue to provide at least a 95%
probability at a
95% confidence level that the hottest fuel rod will not experience Departure from Nucleate Boiling during normal operation or an Anticipated Operational Occurrence initiated from within the LCO limits.
3.3 References for Section 3:
3-1* "Statistical Combination of Uncertainties Part 1",
CEN-123(F)-P
- December, 1979.
3-2* "Statistical Combination of Uncertainties Part 2",
CEN-123(F)-P Harch, 1980.
I 3-3* "Statistical Combination of Uncertainties Part 3",
CEN-123(F)-P February, 1980.
3-4
- Letter, C.
C. Nelson (NRC) to R.
E. Uhrig (FP&L),
"Ammendment ¹48 to Facility Operating License ¹ DPR-67 for St. Lucie 1, Docket ¹50-335,"
November 23, 1981.
These References have been approved for use by the NRC in Reference 3-4.
3-1
TABLE 3-1 STATISTICALLY COMBINED UNCERTAINTIES FOR A DNBR SAFDL OF 1.20 al Approx. values of Equivalent over ower mar in TM/LP LSSS DNB LCO EX-CORE 7.5 8.2 3-2
APPENDIX A A.O A.l MDNBR Probabilit Densit
. Function
~Back round A minimum DNBR probability density function (p.d.f.)
was derived in Reference A-1 in order to arrive at the MDNBR limit of 1.28.
That limit accounted for uncertainties in system
'parameters and provided at least 95% probability and 95%
confidence level that the hot fuel pin would not experience Departure from Nucleate Boiling (DNB).
This MDNBR limit accounted for the following uncertainties:
a) b)
c) d)
e)f) g)
h) i)i) core inlet flow distribution engineering factor on enthalpy rise systematic fuel rod pitch-systematic fuel clad O.D.
engineering factor on heat flux CE-1 Critical Heat Flux (CHF) correlation CE-1 CHF correlation cross validation penalty (5% increase in CHF correlation standard deviation)
T-H code uncertainty penalty (5%, equal to two standard deviations).
fuel rod bow HID fuel design allowance Uncertainties (a) through (h) are the same as those used in
[the extended SCU analysis]
as shown in Table A-1 except for (f).
For the original St.
Lucie Unit 2 SCU analysis, the uncertainty for the CE-1 CHF correlation was based on statistics for the entire CE-1 database, including both CE's 14 x 14 and 16 x 16 fuel.
This distribution had a mean of 0.9998 and standard deviation of 0.0676 and yielded a 1. 13 DNBR limit.
A 0.06 penalty was applied to the SCU DNBR limit obtained with this distribution to accommodate NRC concerns about a more adverse (1. 19)
DNBR limit obtained from the worst subset of data for 16 x 16 fuel.
In the [extended SCU analysis],
the CHF correlation uncertainty is characterized by the p.d.f.
based on data from the subset of 16 x 16 CHF data which yield the 1. 19 DNBR limit approved by NRC.
Uncertainty (i) is a 1.75%
MDNBR allowance to account for rod bow for bundle burnups less than or equal to 30,000 MWD/MTU (Reference A-4).
Uncertainty (j) is an allowance for use of the CE-1 CHF correlation with CE's HID fuel design, which results in a 0.01 DNBR penalty.
A.2 MDNBR PDF for Extended SCU Anal sis The MDNBR p.d.f [used in the extended SCU analysis] is shown in Figure A-l. It includes the system parameter uncertainties discussed in Section A. 1.
However, the uncertainty for CE-1 CHF correlation is based on the distribution (shown in Table A-I) from the subset of data in the CE-1 database for CE's 16 x 16 fuel which yields the 1.19 DNBR limit approved by NRC.
System parameter uncertainties and allowances accommodated in
[the Extended SCU MDNBR pdf] are presented in Table A-1 along with the MONBR mean and standard deviation of the [Extended SCU MDNBR p.d.f.]
A-2
TABLE A-1 Component System Parameter Uncertainties and Allowances Accommodated by [Extended SCU HDNBR pdf]
\\
Com onent Uncertaint hot assembly (channel 9) inlet flow factor channel
[3 8 8*] inlet flow factor channel
[10*] inlet flow factor channel
[16*] inlet flow factor enthalpy rise engineering factor systematic fuel rod pitch (in) systematic fuel clad O.D. (in) heat flux engineering factor CE-1 CHF correlation Thermal Hydraulic code overall HDNBR pdf Hean
.79
.82
.99
.82 1.00
.5055
.3820 1.00
.98".,3**
1.00 1.103 Standard Deviation at 95% Confidence 0.0492 0.0279 0.0147 0.0295 0.015 0.000237 0.000244 0.016 0.08736**
0.03175 0.1080 Channel numbers refer to Fig. A-2, original y presented as Figure 3-5 in Reference A-l.
Statistics based on [1. 19]
DNBR limit for CE's 16 x 16 fuel and including 5% cross validation penalty on standard deviation.
FIGURE A-I DNBR PROBABILITY DISTRIBUTION FUNCTION 4.0 3.5
-(x-p)2 F{x) =
e" 0/2%
202 3.0 0
IM l/)
Ctl Cl 1
CD cC CD CD D
2.5 2.0 1.5 1.0 0.5 0.0 O.B 1.0 J.2
- DNBR, X
CHANNEL NUMBER IN FIRST STAGE MODEL 10 12 13 14 15 16 17 18 19 20 22 23 24 25 I
26 I
I 27 I
I I
I I
2S J
l l
I l
l FiGURE A-2 CHANNEL NUMBERING SCHEME FOR STAGE 1 TORC ANALYSIS TO ESTABLISH RESPONSE SURFACE STATE PARAMETERS A-5
A.3 A-1.
A-2.
A-3.
A-4.
References for A endix A
"Uncertainties Derived by the SCU Methodology, Appendix I to St. Lucie Unit 2 Cycle 2 Reload Safety Report," June 1984.
Letter, J.
R. Miller (NRC) to J.
W. Williams, Jr.
(FP8L),
"Amendment No. 8 to Facility Operating License No.
NPF-16 for the St.
Lucie Plant, Unit No. 2", Docket No. 50-389, November 9,
1984.
- Letter, C. 0.
Thomas (NRC) to A. E. Sherer (CE), "Acceptance for Referencing of Licensing Topical Report CENPD-207 (P/NP)",
C-E Critical Heat Flux:
Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids; Part 2,-Non-uniform Axial Power Distribution", November 2, 1984.
Fuel
& Poison Rod Bowing", June,
- 1983, CENPD-225-P-A.
A-6
B.O Pro osed Chan es to the Technical S ecification Bases B.1 Discussion As presented in the body of this report, the Extended SCU analysis has been performed so that a
[DNB SAFDL of 1.20 with appropriate overpower uncertainties]
assures with a 95%
probability at a
95% confidence level that the hottest fuel pin does not experience departure from nucleate boiling during steady state operations or anticipated operational occurrences.
Therefore,
[the numerical DNBR limit (i.e. 1.20) provides the'5/95 probability/confidence level assurance against DNB occurring when used in conjunction with the appropriate uncertainties calculated]
based on the methods described in this report.
To clarify this point in the Technical Specifications, the proposed Technical Specifications Bases will substitute the numerical value of the DNBR limit (1.28) with the term "the DNB SAFDL Consistent with the methods described in CEN-371(F)-'P".
CEN-371(F)-P is the topical number.
for this report.
Table B-1 presents a list of the affected Technical Specifications Bases.
TABLE B-1 Technical Specifications Bases Requiring Change Due to Incorporation of the ESCU Basis 2.1.1 Basis 2.2.1
-Basis 3/4.2.5 (Reactor Core)
(Thermal Margin/Low Pressure)
(DNB Parameters)
B-2