ML19170A007
| ML19170A007 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/19/2019 |
| From: | Mahesh Chawla Plant Licensing Branch III |
| To: | Exelon Generation Co |
| Chawla M, NRR/DORL/LPLIII, 415-8371 | |
| Shared Package | |
| ML18355A456 | List: |
| References | |
| Download: ML19170A007 (309) | |
Text
B/B-UFSAR
Attachment 4.4A
Additional Information On the Plant Specific Application of t he Westinghouse Improved Thermal Design Procedure To Byron/Braidwood
B/B-UFSAR 4.4A-1 4.4A Additional Infor mation On the Plant Sp ecific Application of the Westinghouse Improved Thermal Design Procedure To Byron/Braidwood The NRC Safety Evaluation Re port on WCAP-9500 entitled Reference Core R eport 17x17 Optimize d Fuel Assembly noted the specific plants using the Westinghouse Improved Thermal Design Proc edure (ITDP) must supply additional informati on on the plant sp ecific application of the ITDP to p erform thermal-hydra ulic analyses.
Thus, Byron/ Bra idwood specific resp onses to NRC information requests a re provided below.
4.4A.1 Request 1 Provide the sensitiv ity factors (S i) and their range of applicability.
4.4A.2 Response 1 The sensitivity factors (S i) and their range of applicability are given in T able 1 of Reference 2 for Byron/Braidwood. Please note that these values are the same as those used in WCAP-9500 with the exception of the range for vessel flow.
The range on flow for Byron/
Braidwood has been exten ded down to 273270 gpm (70%
flow) with no change in the correspond ing sensitivity factor being required.
4.4A.3 Request 2 If the S i values used in the B yron/Braidwood analyses are different than those use d in WCAP-9500, then the applicant must reevaluate th e use of an uncertainty allowance for application of equation 3-2 of WCAP-8567, "Improved Thermal Design Procedure" and the linearity assumption must be validated.
4.4A.4 Response 2 The S i values used in Byron/Br aidwood analys es are the same as those used in WCAP-9500. Therefore, reevaluating the use of an uncertainty a llowance for application of equation 302 of WCAP-8567, "Improved Thermal Design Procedu re" and the li nearity assumption is not required.
4.4A.5 Request 3 Provide and justify the varian ces and distributions for input parameters.
B/B-UFSAR 4.4A-2 4.4A.6 Response 3 The distribution assumed for the input parameters such as pressurizer pressure, core average temperature, reactor power, and R CS flow are normal, two-sided 95+%
probability distributions.
The variances of the se parameters fo r Byron/Braidwood are consistent with the varian ces calculated in the generic response.
Specifically, the uncertainties for pressurizer pressure and cor e average temperature are identical to the generic res ponse since the sensors, process racks, and compu ter and readout devices are standard Westinghouse su pplied NSSS equipment.
Variances in reactor pow er and reactor coolant system flow are calculated based on equation 4 and equation 8 respectively in Refere nce 1. As can be seen from the equations, both prim ary and secondary side parameters are measured for power and flow calorimetrics. The error allowances for the parameters measured by Westinghouse sup plied equipment are identical to those used in the generic submittal (Reference 1). Two input parameters are measu red by non-Westi nghouse supplied instruments. Th ese are feedwater temperature and feedwater pressure. As expe cted, the error allowances for these instruments vary slightly from those used in Reference 1. The error allowances for feedwater temperature and pressure were statistically combined (as described in Reference 1) to get the total channel allowance for each parameter.
The feedwater pressure error allowance w as calculated to be less than the error allowance used in Reference 1.
Therefore, the error contribution to the reactor power and flow uncertainties from fe edwater pressu re is less than that used in the generic response.
Similarly, the error s for feedwater temperature were combined to get the total channel allowance. The total allowance was found to be slightly hig her than that used to calculate RCS flow uncert ainty in Reference 1.
However, the error allowance from feedwater temperature is very small relati ve to the other co ntributing errors and in fact this small additio nal error is absorbed in the statistical comb ination. Therefore, the flow uncertainty calculated in Re ference 1 is applicable for Byron/Braidwood.
As stated in Reference 1, the flow cal orimetric can be performed one of several way
- s. Commonwealth Edison plans to do a precis ion flow calorimetric at the beginning of the cycle and normalize the loop elbow
B/B-UFSAR 4.4A-3 taps. For monthly surve illance to assure plant operation consistent with th e ITDP assum ptions, the loop flows will be read o ff the plant proce ss computer. The total flow uncertain ty associated with this method was calculated in Reference 1 and is applicable to the Byron/Braidwood units.
It is to be noted that the total channel allowance for feedwater temperature was calc ulated to be l ess than the error assumed for the re actor power uncertainty calculation in Reference 1.
Therefore, the power uncertainty for Byro n/Braidwood is bounded by the uncertainty calculated in the generic response.
4.4A.7 Request 4 Justify that the nor mal conditions used in the analyses bound all permitted modes of plant operation.
4.4A.8 Response 4 This item was addressed in R eference 1 and is applicable to the Byron/B raidwood units.
4.4A.9 Request 5 Provide a discussion of what code uncertainties, including their values, are included in the DNBR analyses.
4.4A.10 Response 5 The uncertainties included in the ITDP DNBR analyses for Byron/Braidwood are given in Table 1 of Reference 2. As a result of these values bei ng different from those used in WCAP-9500, the Design DNBR Limits also differ. The calculation of the D esign limit DNBRs for the Typical and Thimble cells are given in Reference 2, Tables 2 and 3 respectively.
Since the Design DNBR Limits given in Table 2 and 3 are different fr om those originally given, Section 4.4 has been rev ised to inco rporate the Reference 2 values.
4.4A.11 Request 6 Provide a block diag ram depicting sens or, processing equipment, computer and readout devices for each parameter channel used in the uncertainty analysis.
Within each element of the b lock diagram identify the accuracy, drift, range, span, operating limits, and setpoints. Identify the ove rall accuracy of each channel transmitter to final output and specify the minimum acceptable accur acy for use with the new procedure. Also identify th e overall accura cy of the
B/B-UFSAR 4.4A-4 final output value and maximum accuracy requirements for each input channel for this final output device.
4.4A.12 Response 6
Block diagrams are n ot provided in t his response.
However, as in the gen eric response, a table is provided in Reference 2 givin g the error breakd own from sensor to computer and readout devices.
This table is abbreviated though, giving only the error breakdowns for instruments that differ from those in Ta ble 4, "Typi cal Instrument Uncertainties," of Ref erence 1. As noted earlier, these instruments are those that measure fee dwater temperature and pressure.
4.4A.13 Request 7
If there are any changes to the THINC-IV correlation, or parameter values outside of previously demonstrated acceptable ranges, t he staff requires a reevaluation of the sensitivity factors and of t he use of equation 3-2 of WCAP-8567.
4.4A.14 Response 7 For Byron/Braidwood, the THINC-IV code and WRB-1 DNB Correlation are the same as that used in WCAP-9500.
Therefore, reevaluat ing the sensitivity factors and the use of equation 3-2 of WCAP-8567 is not required.
References
- 1. Westinghouse letter, NS-EPR-2577, E.
P. Rahe to C. H. Berlinger (NRC), March 31, 1982, proprietary.
- 2. General Electric Compa ny letter transmitting improved thermal design info rmation to the NRC (to be written), proprietary.
B/B-UFSAR 4.6-1 REVISION 7 - DECEMBER 1998 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 4.6.1 Information for Control Rod Drive System (CRDS)
Figure 4.2-8 pro vides the layout of the CRDS. The CRDS is a magnetically operated jack with no hydra ulic system associated with its functioning.
The control r od drive mec hanism consists of four separa te subassemblies.
- a. The pressure vessel which includes the l atch housing and rod travel housings.
- b. The coil stack assembly which includes three operating coils: stationary gripper coil, movable gripper coil and lift coil.
- c. The latch assembly which includes the guide tube, the stationary and the movab le pole piec es and the stationary and movab le gripper latches.
4.6.2 Evaluation of the CRDS The CRDS has been analyzed in detail in a fa ilure mode and effects analysis (Refere nce 1). This study, and the analyses presented in Chapter 15.0, d emonstrates that the CRDS performs its intended safety function, reactor tr ip, by putting the reactor in a subcritic al condition when a safety system setting is approached, with any assumed credible failure of a single active component. T he essential elements of the CRDS (those required to ensure reactor trip) are isolated fr om nonessential portions of the CRDS (the rod control system).
Despite the extremely low probability of a c ommon mode failure impairing the ability of the rea ctor trip system to perform its safety function, analyses have been performe d in accordance with the requirements of WASH-1270.
These analyses, documented in References 2 and 3, have demonst rated that acc eptable safety criteria would not be exceeded even if the C RDS were rendered incapable of functioning during a reactor tran sient for which their function w ould normally be expected.
The design of the control ro d drive mechanism is such that failure of the contr ol rod drive mechanism c ooling system will, in the worst cas e, result in an individu al control r od trip or a full reactor trip.
4.6.3 Testing and Veri fication of the CRDS The CRDS was extensively tested prior to its ope ration. These tests may be subdivided into five catego ries: (1) prototype
B/B-UFSAR 4.6-2 REVISION 8 - DECEMBER 2000 tests of components, (2) prototy pe CRDS tests, (3) production tests of components following manufacture and prior to installation, (4) onsite preoper ational tests, a nd (5) initial startup tests.
In accordance with Table 14.2-65, the reactor trip system operation was verified in a startup test. This test ensured that the system operated in accordance with the safety analysis report, design requireme nts, and plant insta llation. A final test was performed in which a manual reactor trip was initiated, (after fuel load but pri or to initial criticalit y) to verify that all rods would fully insert.
The rod cluster control assembli es were dropped and the drops were timed. The time from b eginning of deca y of stationary gripper coil voltage to dashpot entry shall be less than or equal to 2.7 seconds for each rod, the Technical Specification limit.
In compliance with Tables 14.2
-66 and 14.2-66a, all rods falling outside the two-sigma limit were retested a minimum of three times each. Rods we re dropped into represen tative flow condi-tions. In addition, the CRDS is subject to periodic ins ervice tests.
These tests are conducted to verify the oper ability of the CRDS when called upon to function.
4.6.4 Information for Combined Perfo rmance of Reacti vity Systems As is indicated in Chapt er 15.0, the only po stulated events which assume credit for re activity control systems other than a reactor trip to render t he plant subcritical are the steam line break, feedwater line break, and loss-of-coolant accident. The reactivity control sys tems for which cre dit is taken in these accidents are the reactor trip system and the safety injection system (SIS). Note that no cred it is taken for the boration capabilities of the ch emical and volume cont rol system (CVCS) as a system in the analys is of transients present ed in Chapter 15.0.
The adverse boron dilution possi bilities due to the operation of the CVCS are investigated in C hapter 15.0. Prior proper operation of the CVCS has been presumed as an initial condition to evaluate transients, and appropriate Techni cal Specifications have been prepared to ensure the correct operation or remedial action.
4.6.5 Evaluation of Combined Performance The evaluations of the steam line break, feedwater line break, and the loss-of-coolant accident, which presume the combined actuation of the reactor trip system to the CRDS and the SIS, are presented in Chapter 15.0.
Reactor trip signals and safety injection signals for these events are generated from functionally diverse sensors and actuate diverse means of reactivity control, i.e., control rod insertion and injection of soluble poison.
B/B-UFSAR 4.6-3 REVISION 7 - DECEMBER 1998 Nondiverse but r edundant types of equipm ent are utilized only in the processing of the incoming sensor signals into a ppropriate logic, which ini tiates the protective ac tion. In particular, note that protection from eq uipment failures is provided by redundant equipment and periodic testing. Eff ects of failures of this equipment have been extensively investiga ted as reported in Reference 4. The failure mode and effects ana lysis described in this reference v erifies that any single failure will not have a deleterious effect on the engine ered safety feat ures actuation system.
4.6.6 References
- 1. Shopsky, W. E., "Failu re Mode and Effects Analysis (FMEA) of the Solid State Full Length Rod Control Syst em," WCAP 8976, August 1977.
- 2. "Westinghouse Anticipa ted Transients Without Trip Analysis," WCAP-8330, August 1974.
- 3. Gangloff, W. C.
and Loftus, W. D., "An E valuation of Solid State Logic Reactor Pr otection in Antici pated Transients," WCAP-7706-L (Proprietary) and WCAP-7706 (Nonproprietary), July 1971.
- 4. Eggleston, F. T., Rawlins, D.
H. and Petrow, J. R., "Failure Mode and Effects Analysis (FMEA) of the Engineering Safeguard Features Actua tion System," WCAP-8584 (Proprietary) and WC AP-8760 (Nonpropriet ary), April 1976.
GUIDE THIMBLE FUEL ROD -264 REQ'D OD = 0.360 CLAD THICKNESS
= 0.0225 l..--0.2111
- J 16 SPACES AT 0.1196: 7.936 ___ 8.1118 REF {MID GR IDS) CLAD MATERIAL -ZIRC, ZIRLO OR OPTIMIZED ZIRLO CLADDING REVISION 16 DECEMBER 2016 11 .973 FUEL ASS'Y AND CONTROL ROD DIAGONAL PITCH 8 .1166 TYP FUEL ASSEMBLY WITHOUT ROD CLUSTER CONTROL BYRON/BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-1 17 X 17 VANTAGE 5/VANTAGE
+ FUEL ASSEMBLY CROSS SECTION
REVISION 16 DECEMBER 2016 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-2C 17x17 VANTAGE+ FUEL WITH DEBRIS MITIGATING FEATURES (INCLUDING THE RPG , WIN AND SDFBN COMPONENTS)
REVISION 16 DECEMBER 2016 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-3 WESTINGHOUSE INTEGRAL NOZZLE (WIN)
I . UNASSEMBLED LOCK TUBE ADAPTER PLATE THRU HOLE THIMBLE TUBE REVISION 17 DECEMBER 2018 GRID ASSEMBLED BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-6a TOP GRID TWO BULGE CONNECTION (STARTING WITH BYRON UNIT 2 CYCLE 21 & BRAIDWOOD UNIT 1 CYCLE 21) AND TOP NOllLE ATIACHMENT DETAIL Z l RCCN lUM At.LO't 1HfM9'.t TU!t ------.._
OASH>OT SECTION SfAlNLE:SS sn:tL 90T1'0M GA t D )
STA[Nl.[SS.
su:rt n-H i!fSLE SCRtV fNCONEl PROre:cn VE GfUD st AOa.t:ss sn:n. BOTTOM -----..... t-mZZL.E SPOT !itE:lD INSERT ro CRtO REVISION 14 DECEMBER 2012 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 4.2-7 GUIDE THIMBLE TO BOTTOM GRID AND NOZZLE JOINT
REVISION 15DECEMBER 2014One-Sided 95/95 DNBRToleran a Lunn forDNBR,,S of 1 1311.21.dBYRON/BRAIDWOODSTATIONSUPDATED FINAL SAFETY ANALYSIS REPORTFIGURE 4.4-2bMEASUREDVERSUS PREDICTEDCRITICAL HEAT FLUXABB-NV CORRELATION REVISION 15DECEMBER 20141.2I0.8 ^0.6One-Sided 95/95 DNBRToleranceLimit forDNBR,ysof 1.180,4-0.2 11.41.40.20.40.60,811.2WLOP Predicted CHF, MBtu/6r^ft=BYRON/BRAIDWOODSTATIONSUPDATED FINAL SAFETY ANALYSIS REPORTFIGURE 4.4-2cMEASURED VERSUS PREDICTEDCRITICAL HEAT FLUXWLOP CORRELATION