ML17352A359

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Responds to 931115 RAI Re GL 93-04, Rod Control Sys Failure & Withdrawal of Rod Control Cluster Assemblies.
ML17352A359
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 12/28/1993
From: Plunkett T
FLORIDA POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-04, GL-93-4, L-93-319, NUDOCS 9401070325
Download: ML17352A359 (50)
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(,.ACCELERATED DISTRIBUTION DEMONS ATION SYSTEM IF REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9401070325 DOC.DATE: 93/12/28 NOTARIZED: NO DOCKET g FACIL:50-250 Turkey Point Plant, Unit 3, Florida Power and Light C 05000250 50-251 Turkey Point Plant, Unit 4, Florida Power and Light C 05000251 AUTH. NAME AUTHOR AFFILIATION PLUNKETT,T.F. Florida Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION R Document Control Branch (Document Control Desk)

SUBJECT:

Responds to 931115 RAT re GL 93-04, "Rod Control Sys Failure 6 Withdrawal Of Rod Control Cluster Assemblies." I D

DISTRIBUTION CODE: A030D COPIES RECEIVED:LTR TITLE: Generic Ltr-93-04-Rod Control System I ENCL Failure j SIZE:

6 Wz.thdrawa of Ro Cont S NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL RAGHAVAN,L 1 1 D INTERNAL: NRR/DRCH/HICB 1 1 RR/J3+PW/PDIV-1 1 1 D

NRR/DSSA/SRXB 1 1 EG FILE 01 1 1 EXTERNAL'RC PDR 1 1 D

D D

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WARM CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LIPID FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 6 ENCL 6

0 P.O. Box14000,Juno Beach, Ft. 33408.0420 DEC 28 1993 L-93-319 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Request for Additional Information (RAI)

Generic Letter 93-04, Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies By letter L-93-186, dated August 4, 1993, Florida Power and Light Company (FPL) responded to questions regarding Generic Letter 93-04, Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies. By letter dated Novembex 15, 1993, the NRC requested additional information to support the review of FPL's response to Generic Letter 93-04. The response to the NRC request is enclosed.

Should there be any questions, please contact us.

Very truly yours, T. F. Plunkett Vice President Turkey Point Nuclear Enclosure TFP/CLM/cm cc: S. D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point Nuclear O5Q f'1 v, g PDR P ...

9402070325 931228, ADOCK 05000250

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T F. Plunkett being first duly sworn, deposes and says:

That he is Vice President Turke Point Nuclear, of Florida Power and Light Company> the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

T. F. Plunkett Subscribed and sworn to before me this d'ay of 1993.

ZPA'v~ 4. 8 z~c of Notary Public (Type or Print) v'ame NOTARY PUBLIC, in and for the County of Dade, State of Flori ,:"~'"."p",". GHERYL A. KELLY MY COh%5SION t CC 22$ ?81 BPlRES: eqteeber 27, 199$

My Commission expire C ommissxon No.

T. F. Plunkett is personally known to me.

~ GENERIC LETTER 93-04 RESPONSE TO HRC RAI PAGE I OF 17 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON FPL's RESPONSE TO GENERIC LETTER 93-04 AND THE SALEM ROD CONTROL SYSTEM FAILURES BACKGROUND On May 27, 1993, Salem Unit 2 experienced the uncontrolled withdrawal of a single Rod Cluster Control Assembly (RCCA). The movement of this single RCCA was initially postulated to have resulted from control system logic cabinet card failures (possibly the result of a single initiating failure) coupled with failures and/or effects that had not yet been identified. If this Salem event had been the result of a single failure, the uncontrolled single rod withdrawal event of Hay 27th would have placed the Salem plant outside of its stated FSAR design basis, with the potential for a core power distribution not considered in their original design basis analysis.

As a result of this event, the NRC issued Generic Letter 93-04 (Reference 1),

which requested a written response from Westinghouse licensees under the requirements of 10 CFR 50.54(f). Under Generic Letter 93-04, each licensee was required to provide technical/licensing information to the NRC, which addressed the design basis of the plant with regard to a single failure in the Rod Control System and specified what type of short and long term corrective actions had been taken or were planned for resolution of this issue. In response to Generic Letter'3-04, the design and licensing basis for the Turkey Point Rod Control System was examined and an FPL response was forwarded (Reference 2) based on the best information which was available from Salem at the time.

Upon review of the FPL response to Generic Letter 93-04, the NRC requested additional information as identified in their correspondence of November 15, 1993 (Reference 3). This discussion will serve to clarify FPL's initial response to MRC Generic Letter 93-04 and respond to the NRC Request for Additional Information by providing additional details of the FPL analysis that was performed for uncontrolled asymmetrical control rod withdrawal events.

RESPONSE TO NRC RE UEST FOR ADDITIONAL INFORMATION NRC RE(VESTED ADDITIONAL INFORMATION In response to the FPL response to Generic Letter 93-04, the NRC requested that FPL submit additional information to support its conclusions concerning the Turkey Point design basis for uncontrolled asymmetrical control rod withdrawal events. Specifically, the NRC Staff requested that "In support of your conclusions that you meet the licensing basis for asymmetrical control rod withdrawal events, please provide information and detailed discussions on the application and use of the computer codes, and a comparison of your analysis results with the UFSAR for its applicability, margin to DNBR and validity of analysis for all future cycles."

GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 2 OF 17 FPL RESPONSE 1.0 method of Analysis FPL analyses examined the minimum departure from nucleate boiling ratio (HDNBR),

fuel centerline temperature, reactor coolant system (RCS) pressure and the steam generator pressure to ensure that fuel design safety limits and pressure boundary integrity are maintained during an uncontrolled asymmetric RCCA withdrawal event.

Uncontrolled bank, group, double rod and single rod withdrawals from 100X, BOX, 60X, lOX and hot zero power (HZP). were analyzed using the SIHULATE-3, RETRAN-02 and VIPRE-01 computer codes. 'he loss in DNBR margin was compared to the available margin to ensure that sufficient margin is available to accommo4ate an (

uncontrolled asymmetric RCCA withdrawal event. The peak fuel center line temperature and system pressures were compared to the safety limits to ensure,"

fuel and pressure boundary integrity.

1 The procedure used by FPL in the analysis of uncontrolled asymmetric RCCA withdrawal events is described below:

1 a) Using the SIMULATE-3 physics code, analyze bank, group, double rod and-single rod withdrawals at preselected power levels identifie4 in the FSAR (Reference 23) to calculate the rate of reactivity insertion and the peak FhH. Table I lists the combinations of uncontrolled RCCA withdrawals analyzed.

b) Identify maximum post with4rawal FhH for each of the bank, group, double rod and single rod events. These cases are listed in Table 2.

c) Perform system thermal/hydraulics analyses for the cases identified in step (b) using the RETRAN-02 computer code. Obtain system conditions (core power and inlet temperature) at discrete time intervals up to and including the time of reactor trip.

4) for the cases identified in step (b) re-calculate the power distribution at selected time intervals using the SIHULATE-3 code with the system conditions obtained in step (c). This has the effect of crediting the mitigating effects of reactivity feedback in the calculation of the post withdrawal power distribution.

e) For each case analyzed in step (d), obtain the FdH augmentation factor by dividing the post withdrawal highest FhH by the pre-withdrawal FhH.

f) Normalize all the power distributions using the assumption that at the time of RCCA withdrawal, the FhH in the hottest assembly corresponds to the Technical Specification (Reference 24) Limit. Apply the FhH augmentation factors obtained in step (e) to this limit to yield the values of FAH to be use4 in the ONBR calculation. In addition, a 4X calculational uncertainty is added to the FhH values in the at-power cases. An SX uncertainty is used for the HZP cases.

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' GENERIC LETTER 93-04 RESPONSE TO~NRC RAI PAGE 3 OF 17 g) For all cases, perform DNB analysis of the hottest channel using the VIPRE-01 computer code with the FZN obtained in step (f) and the core conditions at the corresponding time steps obtained from the transient simulations of step (c).

For all cases, compare the minimum DNBR calculated in step (g) for the bank withdrawal to the corresponding minimum DNBR for the group, double rod and single rod uncontrolled withdrawal cases. Calculate the percent change in minimum DNBR for all cases relative to the bank withdrawal.

This change in minimum DNBR represents the DNBR penalty for an asymmetric uncontrolled rod withdrawal relative to the uncontrolled withdrawal of a bank. Tables 3, 4 and 5 list these penalties for asymmetric rod withdrawal from various power levels.

Compare the percent changes obtained in step (h) to the available DNB margin for the uncontrolle'd bank withdrawal events.

The uncontrolled RCCA bank withdrawal analyses described in sections

14. 1. 1 and 14. 1.2 of the UFSAR have been redone by Westinghouse using their latest methodology, the Revised Thermal Design Procedure (RTDP)

(References 4 and 5). The measurement and code uncertainties are statistically combined in this procedure rather than deterministically combined as in the Standard Thermal Design Procedure (STDP) for the analysis of non-LOCA transients.

The RTDP analyses have been completed and gA verified by Westinghouse.

The Westinghouse RTDP methodology has been reviewed and approved by the NRC on a generic basis for application to the analysis of non-LOCA transients in nuclear power plants (Reference 6). However, plant specific analyses for Turkey Point have not been reviewed and approved by the NRC. FPL is scheduled to submit these analyses to the NRC in 1994.

The RTDP analysis results for the uncontrolled RCCA bank withdrawal event for Turkey Point were used as the basis for assessment of the available DNB margin.

Compare the maximum fuel centerline temperature calculated in step (g) to the fuel temperature safety limit of 4800 'F (Reference 5).

Compare the maximum RCS pressure to the design limit of 2750 psia (Reference 5).

Compare the maximum steam generator secondary pressure to the design limit of 1210 psia (Reference 5).

m) The above analyses from 100% reactor power were performed both with the beginning of cycle (BOC) minimum reactivity- feedback and the end of cycle (EOC) maximum reactivity feedback condi'tions. It was recognized that the minimum feedback transients were more severe in terms of the peak reactor power reached during the transient (Table 3). All other power level analyses were performed with minimum reactivity feedback only.

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~ GENERIC LElTER RESPONSE TO NRC 93-04 RAI PAGE 4 OF 17 2.0 Significant Assumptions A number of conservative assumptions are used in the uncontrolled asymmetric RCCA withdrawal analyses. The most significant among these are the following:

i) An FhH corresponding to the Technical Specification Limit is assumed for the fuel pin most affected by an uncontrolled rod withdrawal at the time of event initiation for both OFA and LOPAR fuel assemblies.

Moderator and Doppler reactivity feedback was simulated to represent beginning-of-cycle {BOC) and end-of-cycle {EOC) conditions. The BOC conditions include +5 pcm/ F for moderator temperature coefficient and

-1.0 pcm/ F for Doppler temperature coefficient. The corresponding values for the EOC are 0.0 pcm/ F and -2.9 pcm/ F. The BOC conditions provide minimum negative feedback while the EOC conditions provide maximum negative feedback. These values are conservative relative to the Technical Specification limit of 0.0 pcm/ F to -35 pcm/ F for moderator temperature coefficients at full reactor power.

3.0 Application and Use of Computer Codes

3. 1 SIMULATE-3 Computer Code This is a three dimensional two group (fast and thermal) physics nodal code developed by Studsvik (References 7, 8 and 9), and reviewed and approved by the NRC {References 10 and ll) for reactor physics analysis. The Turkey Point model simulates four quarters of each of its 157 fuel assemblies, each divided into 24 vertical nodes of six inch length. Power distribution in each pin of its 15x15 lattice is inferred from the power calculated for each quarter assembly.

Detailed cross sections are generated by the multi-group transport computer code CASM0-3.

SIMULATE-3 analyses have been compared with plant measured data to demonstrate accuracy of calculation (Reference 12). Attachment 1 to this report provides comparisons of such analyses. SIMULATE-3 predictions agree with the plant measured data within established acceptance criteria described in the attachment.

This code was used to simulate uncontrolled bank, group, double rod and single rod withdrawal from power and hot zero power conditions. Table 1 lists the power levels and the RCCA combinations analyzed in the uncontrolled withdrawal analysis. All of the at-power cases were simulated at the maximum rod withdrawal rate of 72 steps per minute. The manual rate of RCCA withdrawal at the plant is set at 68 steps per minute. The automatic withdrawal rate is variable depending upon the error between T reff and T . However, the plant'is operated with the rod control in manual and the automatic rod withdrawal signal has been from the rod motion controller. The HZP case (uncontrolled bank ave'isconnected withdrawal case only) was simulated with a 75 pcm/sec reactivity insertion rate to match the UFSAR analysis for bank withdrawal. The remaining HIP cases were simulated at the rod withdrawal rate of 72 steps per minute.

GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 5 OF 17 Results from the corresponding RETRAN-02 calculations were used in SIHULATE-3 to vary core inlet temperature and power level to simulate moderator and Doppler feedback as the transient progresses. Results of the SIHULATE-3 analyses for peak FAH and reactivity insertion rates are presented in Table Power distributions calculated by SIMULATE-3 for the uncontrolled RCCA withdrawal cases were provided to VIPRE-01 for DNB analysis.

3.2 RETRAN-02 Computer Code This is a one-dimensional nodal thermal/hydraulics code with a point reactor kinetics model {Reference 13). The code has been developed by EPRI and has been reviewed and approved by the NRC for application to light water reactors

{Reference 14). NRC has also reviewed and approved the use of RETRAN-02 by FPL for the Turkey Point nuclear power plant (Reference 15).

Several benchmark cases were presented to the NRC in support of the above review and subsequent safety evaluation (Reference 16). Bank withdrawal from full power was simulated and compared with the UFSAR results, provided here as Attachment 2.

Reactivity insertion rates corresponding to the bank, group, double rod and single rod withdrawal cases (corresponding to the highest FhH) were simulated using RETRAN-02'. Reactor power increases until the reactor trip occurs either on high nuclear flux or Overtemperature delta T. The analyses were performed both for minimum and maximum moderator and Doppler feedback.

Reactor coolant flow, temperature and pressure calculated by RETRAN-02 were provided to YIPRE-01 For DNB analysis.

Table 3 presents results from the RETRAN-02 analyses of uncontrolled withdrawal of various RCCA combinations at different power levels. Tables 4 and 5 summarize the RETRAN-02 analysis results for the limiting case (which causes the maximum ONBR penalty) which is a single uncontrolled RCCA withdrawal from 60X reactor power with the BOC feedback conditions.

3.3 VIPRE-01 Computer Code VIPRE-OI is a thermal/hydraulics computer code developed by EPRI for DNB analysis and approved by the NRC (References 17 and 18). An eighth core model with forty-nine axial nodes was used for Turkey Point. Westinghouse WRB-I correlation was used for the Optimized Fuel Assemblies {OFA) and W3 L-Grid (W-3L) correlation was used for the Low Parasitic Fuel Assemblies (LOPAR). Turkey Point is in the final stage of transition from LOPAR to OFA fuel.

For the at-power asymmetric RCCA withdrawal cases, VIPRE-Ol'computer code was used to calculate the minimum DNBR, which is the limiting safety criterion. For the HZP cases, the VIPRE-Ol computer code was used to analyze the fuel centerline temperature to ensure that the fuel safety limit was met.

GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 6 OF 17 To ensure confidence in the VIPRE-01 model for Turkey Point, state point values at the time of minimum DNBR in the Mestinghouse RTDP analysis of an uncontrolled bank withdrawal from 60X reactor power were simulated. This analysis has been performed using a reactivity insertion rate of 1 pcm/sec, and the minimum Doppler and moderator reactivity feedback. These state point values are:

Peak reactor power 100X RCS pressure 2350 psia RCS Tave 605 F The minimum ONBR using the M-3L correlation as predicted by VIPRE-01 was 1.80 compared to a value of 1.84 calculated by Mestinghouse. This is considered to be a good agreement.

Table 3 provides results from the VIPRE-01 analysis for the minimum DNBR for various power levels and rod combinations. Tables 4 and 5 summarize the results obtained from the VIPRE-01 analyses for the limiting case (which causes maximum DNBR penalty) which is a single uncontrolled RCCA withdrawal from 60X reactor power with BOC reactivity feedback.

4.0 Uncontrolled RGB Mithdrawal Analysis Results The FSAR uncontrolled RCCA bank withdrawal events have been reanalyzed by Mestinghouse using the RTDP methodology (References 4 and 5). Available margin has been increased as a result of these RTDP analyses. These analyses show that, for the uncontrolled RCCA bank withdrawal event, a minimum of 30X DNBR margin exists for the OFA and 17X DNBR margin exists for the LOPAR fuel assemblies at-power. There are no LOPAR fuel assemblies in the present cycle of Turkey Point Unit 4 and only 5 LOPAR assemblies in the present cycle of Turkey Point Unit 3.

The uncontrolled asymmetric RCCA withdrawal analyses using SIMULATE-3, RETRAN-02 and VIPRE-01 computer codes were performed at 100X, 80X, 60X, 10X and HZP reactor power (References 19 and 20). 'he results of the uncontrolled asymmetric RCCA withdrawal analyses are summarized in Tables 3, 4 and 5. These results show that a single uncontrolled asymmetric RCCA withdrawal results in the maximum DNBR penalty at the BOC feedback conditions with the reactor at 60X power, where the rod movement has a significant impact on both the core power level and distribution. The results for the case of 60X power for OFA and LOPAR fuels, are provided in Tables 3 and 4, respectively. These analyses have shown that the existing DNB margin in the uncontrolled RCCA bank withdrawal analyses for Turkey Point Units 3 and 4, is sufficient to accommodate the DNBR penalty caused by an uncontrolled asymmetric RCCA withdrawal event without violating the DNBR safety limit.

At HZP, an uncontrolled withdrawal of a bank is more limiting than the withdrawal of a group, double rod or a single rod because it results in the maximum peak in the nuclear power. Results of the limiting case of an uncontrolled bank withdrawal from HZP (Table 6), show that the fuel centerline temperatures stay below the safety limit.

GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 7 OF 17 The peak reactor coolant system pressures as listed in Table 3, remain below their safety limit. The peak steam generator secondary pressures, though not listed in Table 3, remain below their safety limit, also.

5.0 Comparison of FPL Analysis Results With the UFSAR for Applicability Attachment 2 provides a comparison of bank withdrawal results from full power.

This transient was analyzed using the RETRAN-02 model of Turkey Point and compared with the FSAR analysis performed by Westinghouse (Reference 16). The analysis results were provided to the NRC for demonstrating FPL capability in the use of RETRAN-02 (Reference 15). This analysis showed good agreement between FPL and Westinghouse analyses in the prediction of reactor power, reactor trip, and RCS temperature and pressure during the transient.

State point comparisons corresponding to the minimum DNBR from a bank withdrawal at 60X power, and corresponding to peak nuclear power from a bank withdrawal at HZP are provided in Table 6. The 60X p'ower case was simulated to compare FPL results from RETRAN-02 and VIPRE-01 with the RTDP analyses performed by Westinghouse. This case assumes a reactivity insertion rate of 1 pcm/sec from a bank withdrawal. There is a good agreement between FPL and Westinghouse predictions of peak reactor power, peak RCS pressure, reactor trip function, reactor trip time, RCS average temperature and minimum DNBR. For this comparison, the moderator temperature coefficient was increased from +5 pcm/sec to +7 pcm/sec to be consistent with the documented Westinghouse analysis.

The HZP case simulated a bank withdrawal with 75 pcm/sec reactivity insertion rate, to be consistent with a documented analysis (Reference 5). The transient is turned around by the Doppler feedback prior to the reactor trip on high neutron flux. The results tabulated in Table 6 show good agreement between peak nuclear power, time of trip, peak heat flux and peak centerline temperature. The system response in this transient is computed using RETRAN-02 and the fuel parameters are computed using VIPRE-01 computer code.

6.0 Validity of analysis for all future cycles Significant parameters affecting the uncontrolled asymmetric RCCA withdrawal analyses are rod worths, core peaking factors, moderator and Doppler temperature coefficients, and axial offset. These parameters will be evaluated for their impact on the uncontrolled asymmetric RCCA withdrawal analyses during the reload safety evaluation process for future fuel cycles. Evaluation of the uncontrolled asymmetric rod withdrawal events for future fuel cycles will continue unless the plant modifications recommended by the Westinghouse Owners Group are implemented to preclude the potential for the uncontrolled rod withdrawal events of the type experienced at Salem.

7.0 Comparison of FPL and Mestinghouse Owners Group Analysis Results for the Uncontrolled Asymmetric RCCA Mithdrawal Event...

Turkey Point Units 3 and 4 were analyzed by the Westinghouse Owners Group (WOG) using RTDP methodology as a representative three loop, 15x15 fuel type plant (References 21 and 22). The WOG results show that the uncontrolled asymmetric rod withdrawal causes a maximum of 5. 1X DNBR penalty for OFA fuel and 8.1X DNBR

~ GENERIC LETTER 33"04 RESPONSE TO NRC RAI PAGE 8 OF 17 penalty for LOPAR fuel, This compares well with the maximum penalties of 7.53X (OFA) and 8.63X (LOPAR) calculated by FPL. Both FPL and Westinghouse analyses show that the existing DNB margin is sufficient to accommodate uncontrolled asymmetric rod withdrawal at Turkey Point.

The Mestinghouse Owners Group report (Reference 21) also states that the results are not significantly dependent on cycle-to-cycle Fuel management changes.

8.0 References NRC Generic Letter 93-04,"Rod Control System Failure and Withdrawal of Rod Cluster Control Assemblies, 10 CFR 50.54(f)," dated June 21, 1993.

2. FPL letter to the NRC L-93-186,"NRC Generic Letter 93 Rod Control System Failure and Withdrawal of Rod Cluster Control Asse'mblies," dated August 4, 1993.
3. NRC letter to FPL,"Turkey Point Units 3 and 4 Generic Letter 93 'Rod Control System Failure and Withdrawal of Rod Cluster Control Assemblies'-

Request for Additional Information (TAC No.s H86873 and H86874)," dated November 15, 1993.

Turkey Point Units 3 and 4,"Non-LOCA Reanalyses for 1990 Fuel Contract,"

Westinghouse 1993.

Turkey Point Units 3 and 4,"Accident Analysis Design Basis .Document,"

Mestinghouse 1993.

6. WCAP-11397-P-A, "Revised Thermal Design Procedure," Mestinghouse, dated April 1989.
7. SIHULATE-3: Advanced Three-Dimensional Two-Group Reactor Analysis Code, Version 3.03, Studsvik, dated April 6, 1990.
8. TABLES-3: Library Preparation Code for SIMULATE-3, Version 3.03, Studsvik, dated April 6, 1990.
9. CASMO-3: Fuel Assembly Burnup Program, Version 4.4, Studsvik, dated November 26, 1990.
10. Ashok C. Thadani (USNRC) letter to Mr. G. Papanic, Jr., Yankee Atomic Electric Company, Acceptance of Referencing of Topical Report YAEC-1659, "SIMULATE-3, Validation and Verification", dated February 20, 1990, Ashok C. Thadani (USNRC) letter to Hr. G. Papanic, Jr., Yankee Atomic Electric Company, Acceptance for Referencing of Topical Report YAEC-1363, "CASHO-3G Validation".
12. FPL Calculation JPN-PSL-OFJF-92-071,"SIMULATE-3 Validation Analysis,"

Revision-l, dated Harch 9, 1993.

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GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 9 OF 17

13. RETRAN-02: A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI NP-1850-CCHA, Vol. 1, Rev. 2, Computer Code Manual, dated November 1984.
14. Cecil 0. Thomas (USNRC) letter to Dr. Thomas W, Schnatz,"Acceptance for Referencing of Licensing Topical Reports EPRI CCH5, RETRAN-A Program for One-Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems, and EPRI NP1850-CCN, RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," dated September 4, 1984.
15. Gus C. Lainas (USNRC) letter to Hr. W. F. Conway, "Florida Power and Light Company - Topical Report on RETRAN (TAC No. 60550) and Topical Report on PWR Physics Hethodology (TAC No. 60549)," dated April 19, 1988.
16. NTH-G-6,"Topical Report, RETRAN Code, Transient Analysis Model gualification," Florida Power and Light Company, dated July 1985.
17. Charles E. Rossi (USNRC) letter to Hr. J. A. Blaisdell,"Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM,"VIPRE-Ol: A Thermal-Hydraulic Analysis Code for Reactor Cores", Volumes 1, 2, 3 and 4, dated Hay 1, 1986.
18. VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, Volumes 1, 2, 3 and 4, EPRI NP-2511-CCM-A, Rev 3, dated August 1989.
19. FPL Calculation JPN-PTN-BFJF-93-041/044,"PTN Physics Data for Rod Withdrawal Analysis at Power and at Hot Zero Power," Rev. 0, dated July 1993;
20. FPL Calculation JPN-PTN-BFJF-93-043," Impact of Asymmetric RCCA Withdrawal on Available DNB Margin for Turkey Point Units 3 and 4," Rev. 0, dated July 1993.
21. WCAP-13803 Rev. 1,"Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," Westinghouse, dated August 1993.
22. Westinghouse Owners Group, Analysis Subcommittee letter OG-93-80,"Revisions to the Results of Asymmetric Rod Withdrawal Analysis Program," dated September 10, 1993.
23. Turkey Point Units 3 and 4 Updated FSAR, Revision 11, dated November 1993.
24. Turkey Point Units 3 and 4 Technical Specifications, Operating License Amendments 157 and 151, effective date November 18, 1993.

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0 GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 11 OF 17 TABLE 1 (Continued)

Description of SIMULATE-3 Cases Evaluated Power Level; Rod Control Rod Cases Full Core Insertion Limit Coordinates (RIL) Steps HZP'RI Bank D Figure 1 Bank C Bank B Bank A Bank SA Bank SB Bank D Group 1 D-8 M-8 Bank D Group 2 H-4 H-8 H-12 Bank C Group assumed 1/4 core Bank B Group symmetry and only Bank A Group evaluated one group Bank SA Group for Banks A-C and Bank SB Gxoup SA, SB Two Rods F-2 G-3 H-4 G-3 F-4;G-3 J-3;G-3 J-5 J-3 H-4;J-3 H-4;K-4 J-3;K-4 Single Rod H-4 H-8 F-2 G-3 D-6 J-3 J-5

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CEXEalC LETTER 93-EESPOgSE 70 KRC RAl pACE 13 Of 17 TABLE 3 RESULTS OF RETRAN-02 AND VIPRE-01 ANALYSES FOR VARIOUS POTfER LEVELS AND RCCA WITHDRAWAL EVENTS CASE PEAK PEAK RCS MDNBR DNB PENALTY ( ) NEW MARGIN DESCRIPTION POWER PRESSURE OFA LOPAR OFA LOPAR OFA LOPAR (MW~h) (PSIA) (WRB-1) (W-3L) (t DNBR) (0 DNBR) 100 0 Power Min. Feedback-Bank 2500.6 2306.9 2.06 1.72 Group 2436.9 2308.9 2 09 1.76 -1.56 2 ~ 27 30. 63 19. 67 Two Rods 2.04 1.75 0.68 -1.63 28.39 19.03 Single Rod 2398.3 2309.4 2.08 1.76 -1.12 -2.56 30.19 19.96 Max. Feedback (3)

Bank 2397.0 2309.3 Group 2305.5 2278.0 Two Rods Single Rod 2255.7 2252.0 80  % Power :

Min. Feedback Bank 2461.8 2335.4 1.95 1.45 Group 2321.1 2326.8 1.92 1.48 l. 49 -2.28 27.58 19. 68 Two Rods 1.80 1.44 7.44 0.21 21.63 17. 19 Single Rod 2253.5 2343.6 1.84 1.47 5.39 -1.80 23.68 19.20 60  % Power Min. Feedback Bank 2312.4 2352.3 1. 98 1. 80 Group 2174.7 2338.9 1.95 1. 75 1.21 2.39 27.86 15.01 Two Rods 1.89 1.70 4.60 5.46 24.47 11.94 Single Rod 2076.0 2341.2 1.83 1.64 7.53 8.63 21.54 8.77 10  % Power Min. Feedback Bank 2215.1 2352.5 1. 72 l. 56 -12.40 -7.56 Group 1571.5 2347.0 1.93 1. 68 41.47 24.96 Two Rods 1693.3 2334.8 1. 61 1.43 6. 46 8.59 22.61 8.81 Single Rod 1509.1 2338.9 1 75 1.53 -1.92 1 99 30.99 15.41 HZP Power Min. Feedback Bank 0. 66 2229.1 1. 45 l. 36 5.50 Group 0.21 2258.3 5 ~ 88 Two Rods 0.43(2) 2247.2 4.13 3.78 Single Rod 0.21(2) 2268.3 3.57 3.20 NOTES 1~ For all cases initiated from hot zero power (HZP), the MDNBR calculated for the bank withdrawal is limiting. Therefore, asymmetrical withdrawal cases from hot zero power are bounded by the reference bank withdrawal from hot zero power for DNB considerations.

2. Maximum core average heat flux as a fraction of nominal.
3. Since minimum feedback resulted in higher power peaks at 100'L power, all other cases were analyzed with minimum feedback only.

~ GENERIC LETTER RESPONSE TO NRC 93-04 RAI PAGE 14 OF 17 TABLE 4 ROD WITHDRAWAL FROM 60t POWER (BOC)

OFA FUEL PARAMETER BANK GROUP DOUBLE SXNGLE W/DRAW W/DRAW ROD ROD W/DRAW W/DRAW REACTXVXTY XNSERTION 6.2 7 2.6 1.4 RATE, AVERAGE {pcm/sec)

TXME OF REACTOR 41 5~ 74. 1 74 ~ 1 106 '

TRIP (sec)

TRIP SIGNAL OTAT OThT OThT OThT

  • FhH REACTOR TRXP 1.82 1.88 1.93 2 '3 MINIMUM DNBR 1.98 1 ~ 95 1. 89 1.83 DNBR PENALTY RELATXVE TO BANK WITHDRAWAL (4)
0. l. 21 4. 60 7 '3 REMAINING MARGXN TO 29 ' 27.79 24.40 21. 47 DNBR LIMXT (4)

NOTES

  • OThT = Over Temperature delta T Trip

GENERIC LETTER 93-04 RESPONSE TO NRC RAI PAGE 15 OF 17 TABLE 5 ROD WITHDRAWAL FROM 60% POWER (BOO)

LOPAR FUEL PARAMETER BANK GROUP DOUBLE SINKS

'W/DRAM W/DRAW ROD ROD W/DRAW, W/QQR REACTIVXTY XNSERTION 6.2 ' 2.6 1.4 RATE, AVERAGE (pcm/sec)

TXME OF REACTOR 41 5~ 74,1 74. 1 106.3 TRIP (sec)

TRIP SIGNAL OTAT OTAT OTAT OTAT FhH REACTOR TRIP 1.82 1 '8 1.93 2 '3 MINIMUM DNBR 1.80 1. 75 1.70 1 ~ 64 DNBR PENALTY RELATIVE TO BANK WXTHDRAWAL (4')

0. 2 '9 5 46 8.63 REMAXNXNG MARGIN TO 17 ' 14 ~ 61 11. 54 8.37

Notes:

  • Additional margin is present in the analyses because the latest RTDP analyses (relative to which the above margins are quoted) assumed the following conservatism:
1) 204 steam generator tube plugging (presently less than 54 tubes are plugged),
2) the LOPAR fuel assemblies are at least once-burned and thus run well below the allowable FAH limit.

There are no LOPAR fuel assemblies in Unit 4 and only 5 assemblies in Unit 3 in the present, cores.

GEMERIC LETTfR 93-04 .

RESPONSE TO NRC RAI PAGE 16 OF 17 TABLE 6 FPL VERSUS WESTINGHOUSE (RTDP) BANK WITHDRAWAL COMPARISON RCCA BANK WITHDRAWAL FROM 604 POWER (1 PCM/SEC)

P2QUQTETER WESTINGHOUSE Peak Power (4 Nominal) 1004 934'318 Peak RCS Pressure 2350 psia psia Core Tave -9 Reactor Trip 605 F 594 F Trip Function OTb,T OTLT Time of Trip 104.1 sec 100.7 sec Minimum DNBR (W-3L) 1.84 1 '7 RCCA BANK WITHDRAWAL FROM HOT SERO POWER (75 PCM/SEC)

PARAMETER WESTINGHOUSE FPL Peak Nuclear Power (4 Nominal) 2004 263%

Peak Heat Flux (4 Nominal) 554 744 Trip Function Hi Flux (354) Hi Flux (35%)

Time of Trip 10.3 sec 11.4 sec Peak Center-line Temperature 2538 F 2486 oF Peak Fuel Average Temperature 2148 F 2106 F Peak Clad Temperature 725 F 757 F

GENERIC LETTER 93-04 RES PONS E TO NRC RA I PAGE 17 OF 17 FIGURE 1 RPNMLK JHG F.E DC Bq I

2 5

6 SA 7 D Sg D Sp SA SA 9 10 Sp II D 12 13 ABSORBER MATERIAL:

Ag-In-Cd ld CONTROL ROO OESIGNATION 15 FUNCTION NUMBER OF CLUSTERS CONTROL BANK 0 5 CONTROL BANK C 8 CONTROL BANK B 8 CONTROL BANK A 8 SHUTDOWN BANK S 8 SHUTDOWN BANK S 8 CONTROL ANO SHUTDOWN ROD LOCATIONS

Attachment 1 Page 1 of 12 Validation of FPL SIMULATE-3 Core Physics Models

Reference:

JPN Calculation PSL-OFJF-92-071, Revision 0, "SIMULATE-3 Validation Analysis," Approved 3/9/93 The SIMULATE-3 core physics models have been validated against measured data. Included here is part of the validation performed consisting of axial power shapes, control rod worth comparisonsg and radial power distributions for various units and cycles. The good comparisons between measured and calculated radial power distributions, axial power shapes, and control rod worths provides the justification for the use of SIMULATE-3 to calculate control rod insertion rates and peaking factors used in the uncontrolled rod withdrawal analyses.

The following acceptance criteria were used during these benchmarks. These criteria were obtained from ANSI 19.6, Technical Specifications, Operating Procedures and Industry experience.

Radial power distribution +0.100 for each measured assembly power & RMS <5%

Axial power distribution +0.03 axial offset units Control rod worths Individual banks +15% or +100 pcm whichever is greater Summary of Attached Comparisons Descri tion Pacae Axial Power Shape: Turkey Point 3 cycle 11 2 Turkey Point 3 cycle 12 3 Turkey Point 4 Cycle 13 4 Control Rod Worth: Turkey Point 3 cycle 10 Turkey Point 3 cycle 11 Turkey Point 3 Cycle 12 Turkey Point 4 cycle 11 Turkey Point 4 cycle 12 Turkey Point 4 Cycle 13 Radial Power Turkey Point 3 BOC 12 6 Distribution: Turkey Point 3 MOC 12 7 Turkey Point 3 EOC 12 8 Turkey Point 4 MOC 12 9 Turkey Point 4 EOC 12 10 Turkey Point 4 BOC 13 11 Turkey Point 4 MOC 13 12

ATTACHMENT 1 page 2 of 12 Table 6.3e Axial Shape Index Comparisons Turkey Point Unit 3/Cycle 11

[1) (2) (3)

BURNUP SIM4 MEAS FLUXMAP FLUXMAP DIFF. VENDOR DIFF (GWD/MTU) AO AO ID DATE (S3-M) AO (S3-VEND) 1.080 0.033 0.042 FM3XI6 03/15/88 -0.009 0.004 0.029 1.700 0.023 0.034 FM3XI8 04/19/88 -0.011 -0.003 0.026 2.545 0.011 0.024 FM3XI9 05/17/88 -0.013 -0.012 0.023 3.260 0.003 0.015 FM3XI10 06/09/88 -0.012 -0.018 0.021 4.075 -0.005 0.007 FM3XI11 07/07/88 -0.012 -0.024 0.019 4.875 -0.011 0.003 FM3XI12 08/03/88 -0.014 -0.027 0.016 5.746 -0.015 -0.004 FM3XI13B 09/06/88 -0.011 -0.029 0.014 6.249 -0.014 -0.005 FM3XI14 09/16/88 -0.009 -0.030 0.016 7.710 -0.018 -0.002 FM3XI16 03/16/89 -0.016 -0.033 0.015 8.487 -0.018 -0.004 FM3XI17 07/10/89 -0.014 -0.033 0.015 9.196 -0.019 ~ 0.038 FM3XI18 08/03/89 0.019 -0.032 0.013 9.752 -0.019 -0.013 FM3XI19R 08/23/89 -0.006 -0.032 0.013 10.485 -0.019 -0.021 FM3XI20 09/19/89 0.002 -0.031 0.012 11.140 -0.019 -0.016 FM3XI21 10/10/89 -0.004 -0.030 0.011 11.864 -0.019 -0.019 FM3XI22 11/06/89 0.000 -0.029 0.010 12.735 -0.019 -0.022 FM3XI23 12/05/89 0.003 -0.029 0.010 13.570 -0.019 -0.024 FM3XI24 01/03/90 0.005 -0.030 0.011 AVERAGE DIFFERENCE: -0.015 0.016 STANDARD DEVIATION: 0.0135 0.0055 Notes:

1. SIM-3 rcfcrcocc j9387, lll483.
2. Mcasorcd data frosts INCORE-3D, vcr.3.$ fioanaps.
3. Vendor data from Wcttirgtoousc WCAP-11454, 4/37, PCNDR.

TP3 CYCLE I I CORE AVERAGE AO I I om Om 0 OS) 1 1

I

  • I I

1stto $ 5l5 assr 5 5744 7710 9494 t 0845 11 444 1 5470 1.700 $ 260 4 $ 15 Burnup (GWD/MTU)

S3V303 ~ INCORE38 ~ WESTINGHOUSE (SIIIULATE-3)

ATTACHMENT l Page 3 of l2 Table 5.3f Axial Shape Index Comparisons Turkey Point Unit 3/Cycle 12 (ll (2l (3)

BURNUP SIM-3 MEAS FLUXMAP FLUXMAP DIFF. VENDOR DIFF (GWD/MTU) AO AO ID DATE (S3-M) AO (S3-VEND) 782 0.012 0.010 FM3XII04 07/11/90 0.0024 0.006 0.0056 1552 0.009 0.001 FM3XI105 08/06/90 0.0080 -0.001 0.0103 3,140 -0.002 0.013 FM3XII07 09/28/90 -0.0146 -0.013 0.0113 4,740 -0.010 0.005 FM3XII09 11/20/90 -0.0150 -0.021 0.0105 5,646 ~ 0.014 0.001 FM3XII12 10/17/91 -0.0154 -0.021 0.0072 7,222 -0.016 -0.004 FM3XII14 12/09/91 -0.0118 -0.022 0.0061 8,945 -0.017 ~

0.010 FM3XII16 02/05/92 -0.0073 -0.022 0.0047 9,743 -0.015 -0.016 FM3XII17 03/06/92 0.0005 -0.021 0.0062 11,026 -0.015 -0.012 FM3XII19 04/23/92 ~

0.0032 -0.021 0.0058 12,222 -0.016 ~ 0.009 FM3XII21 06/22/92 -0.0072 -0.021 0.0053 13,050 -0.016 -0.004 FM3XII22 07/23/92 -0.0123 -0.021 0.0055 AVERAGE DIFFERENCE: -0.0069 0.0071 STANDARD DEVIATION: 0,0076 0.0023 Xotcst

l. SIM-3 data from j9244. 10/29/92.
2. Measured data for fluxmaps prior to FM3X1119 are from lYCORE.3D, ver. 3.8 calculations.

Data for FM3X1119 and lata wac calculated using lYCORE-3D, va. 7.2.

3. Vendor data from Westinghouse WCAP-12538, 4/90, Fig. 3-16.

TP3 CYCLE 12 CORE AVERAGE AO I I 1

I 0.01 4.01 I

I 4 I I

I I l

I 1

I i i  !

782 1S52 S.tso 4.740 5.646 7.~ 8.945 9.743 11.026 12.222 13.0SO Burnup (GWD/MTU)

(sIIIULATE-3) S3V303 ~ INCORE38

~ INCORE72 .WESTINGHOUSE

ATTACHMENT 1 page 4 of 12 Table 5.3g Axial Shape index Comparisons Turkey Point Unit 4/Cycle 13 (21 (3)

BURNUP SIM-3 MEAS FLUXMAP FLUXMAP DIFF. VENDOR DIFF (GWD/MTU) AS I ASI ID DATE (S3-M) AS I (S3-VEND) 1536 0.022 0.024 FM4135 12/31/91 -0.0018 0.017 0.0048 3224 0.003 0.009 FM4137 03/04/92 -0.0061 -0.006 0.0092 4031 -0.003 0.001 FM4138 03/30/92 -0.0041 -0.012 0.0093 4888 -0.009 -0.004 FM4139 04/27/92 -0.0045 -0.0205 0.0119 5788 -0.013 ~

0.004 FM41310 05/26/92 0.0094 -0.0208 0.0078 6678 .0.016 -0.007 FM41311 06/25/92 -0.0094 -0.023 0.0070 7,461 -0.018 -0.012 FM41312 07/21/92 -0.0062 -0.0245 0.0062 8,265 -0.019 -0.015 FM41313 08/19/92 -0.0042 -0.0251 0.0058 AVERAGE DIFFERENCE: -0.0057 0.0078 STANDARD DEVIATION: 0.0025 0,0022 Notes:

I. SIM-3 reference j8357, 9/28/92.

2. Vendor dsts from Westinghouse WCAP-13021. Si91, Fig. 3.18.

ops TP4 CYCLE 13 CORE AVERAGE AO OAt O o r I I I ts)4 s~ 40st 4ss srss ssrs 7,45l s~

Bumup (GWD/MTU)

S3V303 INCORE72 a WESTINGHOUSE (SIHULATE-3)

ATTACHMENT 1 page 5 of 12 Table 5.4 Control Rod %orth Comparisons TURKEY POINT 3/CYCLE 10 TURKEY POINT 4/CYCLE 11 A(LL AOJ.

BA)8( MEAS SIM4 DIFFERENCE BANK MEAS SILL3 DIFFERENCE IO (PCM) (PCLI) (PCM) (Yi) IO (PCM) (PCM) (PCM) (%)

0 654 M3 9 'I 44Yi 0 771 753 ~ 18 ~ 2,3!Yi C 1414 1419 5 0.36% C I 532 1458 .76 4.97Yi 8 515 553 38 7.32Y 8 607 632 24 3.94Yi A 1170 1171 I 0.09Yi A 1161 1084 ~ 77  %.64Yi SB 1091 1057 44 SB 1164 1110 ~ 54 4.66Yi SA 1083 1125 42 SA 1249 1154 ~ 95 ~ 7.6332 TOTAL 5927 5988 61 1 03Yi TOTAL 6485 6188 .297 A.STY AVE AVE STD OEV STD OEV 41 Sits L45-52C. I III A5 Sih lA&477 ll/l1h4 554 5)5 II I. IOOIh2 l 5DI jl IOL!/Ithl SETA ASST lh54 SETA AllLm TURKEY POINT 3/CYCLE 11 TURKEY POINT 4/CYCLE 12 ADJ. ADJ.

BANK MEAS SIL14 DIFFERENCE BANK MEAS SILS0 DIFFERENCE IO (PCM) (PCM) (PCM) (Yi) IO (PCM) (PCLI) (PCM) ('A) 0 694 635 .59 4.54Yi D 708 712 4 0.52Y C 1349 1310 39 ~ 2.88% C 1347 1339 4.ae 8 632 661 29 4.56Yi 8 384 413 29 7.44Yi A 1109 969 .140 -12.59Y A 1206 1157 A9 A.ICY SB 1127 1017 ~ 110 ~ 9.75Y SB 1210 1198 ~ 12 4.95Y SA 1070 1058 ~ 14 I.30% SA 1025 1043 18 1.75Yi TOTAL 5981 5648 .333 .5.56% TOTAL 5881 5862 19 432!L AVE ~ SS ~ $ .0516 AVE ~ 3 0.65%

STD OEV 57 SA1 Yi STD OEV 25 341 Yi Sa/I L4$ .104 5OlhS l4ck L It-IOt. 4/21ht 5SI 5/tIXI. I/IIhl 5121 I I225L I/Ith)

SSTA ADI> ISIS SSTA ADLi TURKEY POINT 3/CYCLE 12 TURKEY POINT 4/CYCLE 13 ADJ. AOJ.

BANK LlEAS SIM4 DIFFERENCE BANK MEAS SIM4 DIFFERENCE IO (PCM) (PCM) (PCM) (Yi) IO (PCM) (PCM) (PCM) (Yi) 0 856 821 45  %.11Yi 0 661 664 3 0.52Y C 1389 1435 45 3.30Y C 'I 052 1046 4 4.58!L 8 457 479 22 4.82Y 8 448 'l8 3.95Yi A 1143 1153 10 0.88Yi A !270 1236 34 ~ 2.65Yi SB !038 1084 45 4.47Yi SB 1219 1175 ~ 3.62Y SA 1217 1206 ~ 11 4.90Y SA 851 871 20 2 33Yi TOTAL 6100 6178 78 1.28Yi TOTAL S501 5458 23 428Yi AVE 13 1.41 Yi AVE ~ 7 4.01Yi STD OEV 29 3.18Yi STD OEV 24 2.64Yi Si/I L40 2tt. I/IIhO Sil: L 224CO. I/25h2 5ISI 515221. IGOIh2 5!M 5 IIIOI.I/25h2 SSTA A OI> I'll SETA ADI~ I.OII

ATTACHMENT 1 Page 6 of l2 TURKEY POINT UNIT 3. CYCLE 12 RELATIVE POWER DENSITY 1.552 GWD/HT ( 1225 EFPH) 2 3 4 5 0 7 3 1.024 1.303 1. 281 ) . 003 1.315 ).309 0 '73 0.245

).029 1.308 1.274 0.990 1.287 1.277 ' '80 0.258

-0.005 -0.005 0.007 0.013 0.028 0 032 0 007 -0 013 F

~

z02 ).102 1.223 1.352 1.241 1.279 1.062 0.228 09 1.120 1.213 1.315 1.211 1.247 1.062 0.242

-0.007 -O.O)8 O.O)O O.O37 0.030 0.032 0.000 I -0.0)4

%%%%W}f%

\ o8) ).221 % 1.358  % 1.214 1.225 1.251 0.769 3 ).293 1.239 w 1.348  % 1.196 1.200 1 ~ 210 0.792

-0.012 0.025 0.041 -0.023 0.998 1.343 )f 1.212 ).029 1.276 0.981 0.341 1.021 w 1.368 % 1.221 l. 042 1.246 0.969 0.341 O.D23 %-0.025 %-0.009 -O. 013 0.030 0.012 0.000

).303 1.223 1.2)S ).267 0.837 0.380 5 1.322 1.262 1.244 1. 264 0.839 0.377

-0.019 -0.039 -0.026 0.003 -0.002 0.003 1.302 1.270 1.242 0.971 0.373 6 1.321 1.297 1.252 0.973 0.381

-0.019 -0.027 -0.010 -0.002 -0.008 0.765 ).055 0.764 0.339 MIN. > -0, 039 MAX. = 0.041 0.780 1.D68 0.775 0.341 R.H.S.

  • 0.020

-0.015 -0.013 -0.011 -0.002 0 '42 0.226 X.XXX SIHV303 928 PPH 8 0.250 0.236 X.XXX FH3XII5 0 PPH

-0.008 -0.010 X.XXX DIFFERENCE CH2.SVVR.TP312.J2222.OUTPUT JDB < VFRXTJCS . J02222 - 24 SEP 92

ATTACHMENT 1 Page 7 of 12 TURKEY POINT UNIT 3, CYCLE 12 RELATIVE POlJER DENSITY 7.222 GND/NT ( 5700 EFPN) 1 2 3 5 6 1.062 < 1.392 x 1.215 0.963 1.212 1.327 0.766 0.260 1 1.070 w 1.410 % 1.243 0.976 1.224i 1.335 '.761 0.255

-0.008 %-0.018 3E-0 '28 -0.013 -0.012 -0.008 0.005 0.005 x 1.393 ~ 1.084 1.165 1.373 1. 153 1.198 1.048 0 '42 Zx 1.403 w 1.113 1.176 1.337 1.151 1.201 1.034 0.235

+-0.010 >-0.029 -0.011 0.036 0.002 -0.003 0.014 0.007 1.165 1.373 1.14i8 1.161 1.290 0.783 1.243 1.188 1.351 1.139 1.151 1.265 0.762 0 '24 -0.023 0.022 0.009 0.010 0.025 0. 021 0.964 1.371 1.148 0.993 1.299 1 ~ 117 0.387 0.980 1,373 1 ~ 14i8 0.981 1.257 1.068 0.370 0 F 016 -0 '02 0.000 0.012 0.042 0.049 0.017

1. 211 1. 145 1 ~ 159 1.294 0.885 0.433
1. 231 1. 169 1.164 1.269 0.865 0.421

-0.020 -0.024 -0.005 0.025 0.020 0.012 1.330 1.199 1.289 1, 112 0. ri28

1. 325 1. 210 1. 272 1.089 0.425 0.005 -0 011 F 0.017 0,023 0 F 003 0.765 1.049 0.782 0.387 NIN. = -0. 029 MAX. R 0.049 0.771 1 ~ 054 0.770 0.381 R.N.S. ~ 0.019

-0.006 -0.005 0.012 0.006 0.260 0.24i2 X.XXX 5INV303 553 PPM 0.266 0.248 X.XXX TPSXII14 0 PPN

-0.006 -0.006 X.XXX DIFFERENCE CN2.SVVR.TP312.J2222.OUTPUT JOB ~ UFRXTJCS - J02222 - 24 SEP 92

ATTACHMENT l Page 8 of l2 TURKEY POINT UNIT 3, CYCLE 12 RELATIVE POMER DENSITY 12.222 GQD/ HT ( 96ii7 EFPN) 2 3 5 6 7

).032 1.333 X 1.170 0.953 1.176 1.302 0.79ii 0. 29ii 1 1.007 1.356 5 1.185 0.951 1.120 1.329 , 0.797 0.293 0.025 %-0.023 %"0.0)5 0.002 0.056 -0.027 -0.003 0.001

).334 1.052 1.125 1.326 1. ) 2ii 1.176 1 '65 0.272 2 ).3ii3 1.083 ).)ii3 1.327 1. 120 ).)Bi'i 1.082 0.267

-0.009 -0.031 -0.018 -0.001 0. 00ii -0.008 -0.017 0.005 1.173 1.125 1.336 w 1.123 ).)ii2 1.291 0 '19 3 1. 17ii 1. ) Ii0 1.335  % 1.121 1 ~ 126 1.298 0.816

-0.001 -0.015 0.001  % 0.002 0.016 -0.007 0.003 0.955 1.326 ).12ii 0.999 1.300 1.169 0 Ii3ii

~

ii 0.935 1.317 1.132 1. 00i'i 1.309 1. 163 0. ii26 0.020 0.009 -0.008 -0.005 -0.009 0.006 0.008

1. 176 1. 118 1. ) Ij2 1. 297 0.937 0. ci86 5 1.122 1. 086 1. 150 1. 3ii2 0.962 0. Ij82 0.05ci 0.032 -0. 008 -0. Oii5 -0.005 0.004 1.306 1.178 1. 292 1. 166 O.482 6 ).303 1.182 1.317 1.187 0. ii75 0.003 -0.00ii -0 '25 -0.021 0.007 0.794 1.066 0.820 O.ii3ii HIN. ~ -0.045 HAX. > 0.056 0.783 1.086 0.821 0. Ii2Ii R.H.S. > 0.017 0.011 -0.020 "0.001 0.010 0.295 0.273 X.XXX SIHV303 88 PPH 8 0.292 0 '6ii X.XXX FH3XII21 0 PPH 0.003 0.009 X.XXX DIFFERENCE CN2.SVVR.TP312.J2222.OUTPUT JOB
  • UFRXTJCS - J02222 - 2ii SEP 92

ATTACHMENT 1 page 9 of 12 TURKEY POINT UNIT 6p CYCLE 12 RELATIVE POWER DENSITY 7.620 GWD/MT ( 5999 EFPH)

+-------+

1.117 1.152

-0.035

+ +

1.071 1.364 I I

1. 095 1.393 i
-0.024 -0.029

'.'68 I 1.165 1.059 I 1.181 1.182 1.076

!-0.013 -0.017 -0.015

+

i l.'65 118 l 1. 376 1.125 I

i 1.171 116 '357 1.122

'-0 006 0.002 0 '19 0.003

+ +

1.064  ! 1.373  ! 1.135 I 1.170 1.156 i 1.072 '372 I 1.136 ,'.166 I

1.130 I I I

!-0.028 0.001 !-0.001 I 0.006 0.026

'+ +

'338 '.080  ! 1.291 I I

1.135 0.671 I

1-. 358 1.111 1.272 l 1.109  ! 0.653  !

.-0.020 -0.031 0.019 I 0.026 0.018

'r 0 820 ' 048 'I

0. 679 0.388 MIN. = -0.035 I MAX. = 0.026 0.8ci9 1. 051 0.660 0.373 R.M.S. = 0.018 I

-0.029 -0.003 0.019 0.015

+-------+

+ '+

0.296 I 0.258 I X.XXX SIM3V303 587 PPM 0.301 I 0.249 i X. XXX FMQXII12 I 0 ~PM

"'-0.005 ' '09 +--- XXX:

X

- + DIFFERENCE JOB = UFRXTJCL - J08126 - 15 JAN 93

ATTACHMENT 1 Page 10 of 12 TURKEY POINT UNIT 4, CYCLE 12 RELATIVE POWER DENSITY 11.812 GWD/HT ( 9299 EFPH)

+ +

1.108 1.147

-0.039

+

l 1.065 1.356 I

1.091 1.390 i I

!-0.026 l-0.034

+ +

1.149 1.148 1.056 1.158 1.174 1.075

-0.009 -0.026 -0.019 I. + + +

1.148 1. 106 0 372 l. 113 160 1.122 1.359 '.. 104

'-0 0'2 -0.016 0.013 0.009

+ +

l 1.040 1.367 1.121 1.155 1.164 I

1.073 1.361 1 ~ 113 1.145 1.146 I

l-0.033 0.006 0.008 0 F 010 0.018

'+ +

l 1.315 1.070 1.274 1.143 0.499 I

1.342 1.055 1.252 '.119 0.475 I

l-0.027 0.015 0.022 I 0.024 0.024

+

I 0.836 1.054 0. 701 l 0. 415 MIN. = -0. 039 I HAX. = 0.027

! 0.853 1.044 0.674 0.394 R.H.S. = 0.019 I

-0.017 0. 010 I 0. 027 0. 021

+-------+

0.323 0.281 l X.XXX SIH3V303 198 PPH I 0.327 l 0.269  ! X.XXX FH4XII18 l I 0 PPH

-0.004 0.012 X.XXX DIFFERENCE

+

JOB = UFRXTJC' J08126 " 15 JAN 93

ATTACHMENT l Page ll of 12 TURKEY POINT UNIT 4, CYCLE 13 RELATIVE PONER DENSITY 0.919 GMD/HT ( 724 EFPH) 3 5 1.044 1 1. 0444 1.368 If 1.082

'If 2)f 1.330 If 1.063 i

0.038 If 0.019

ÃIIKNlf%%MMM%%%%%%K 1.089 If 1.363 If 1.290 3 1.089 If 1.343 % 1.283 0.000 If 0.020 If 0.007 1.308 1.257 1.24i6 1.304 1.278 1.257 1.255 1. 320 0.030 0.000 -0.009 -0.016 0.979 1.281 0.973 1.110 1.053 1.017 1.283 1.009 1.129 1.039

-0.038 -0.002 -0.036 -0.019 0.014 1.315 0.968 1.250 1.087 0.435 1.280 0.991 1.230 1. 023 0. 431 0.035 -0.023 0.020 0.064 0.004 0 907 i 1 '20 I I

0 788 0.397 MIN.

MAX.'

> -0. 038 0 '64 7 0.917 4 20 0.770 0.393 R.H.S. > 0.023

-0.010 0.000 0.018 0.004 0.263 0.236 X.XXX S3V303 1055 PPH 0.273 0.251 X.XXX TP4XIII4 0 PPH

-0.010 -0.015 X.XXX DIFFERENCE CM2.SVVR.TP413.J93443.OUTPUT JOB > UFRNRJR4 - J09343 - 24 SEP 92

ATTACHMENT 1 Page 12 of 12 TURKEY POINT UNIT 4, CYCLE 13 RELATIVE POWER DENSITY 7.461 GMD/NT ( 5875 EFPH) 5 1.018 1 1.023

-0.005 1.364 1.016 2 1.346 1. 000 0.018 0. 016 1.037 1.238 1. 14i8 1 ~ 033 1.218 1. 138 0.004 0.020 0.010 1 ~ 387  % 1.196 1.143 1.207 4% 1.392  % 1.197 1. 135 1. 179

%-0 '05  %-0.001 0. DDB 0.028 1.016  % 1.361  % 0.970 1.098 1.173 1.059  % 1.399  % 0.998 l. 081 1.147

-0.043 %-0.038  %-0.028 0.017 0.026 1 ~ 34i3 0.984 1.304 1.151 0 ~ 494 1.387 1.015 1.316 1. 138 0 ~ 494

-0.044 -0.031 -0. 012 O.D13 0.000 0.905 l.'09 0.813 0.440 HIN. * -0.044 0 ~ 902 1 . 124 0.833 0.456 HAX.

R.H. S. ii 0.028

0. 021 0.003 -0.015 -0.020 -0.016 0.285 0.255 X.XXX S3V303 534i PPH 0.289 0.267 X.XXX TP4XI II 12 0 PPH

-0.004 -0.012 X.XXX DIFFERENCE JOB i CH2.SVVR.TP413.J9343.OUTPUT UFRNRJR4 - J09343 - 24 SEP 92

Var,

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ATTACHMENT 2 Page X of 8 6.0 REACTIVITY INSERTION Events in this category involve localized reactivity additions which cause anomalies in the core power distribution. Important modeling considerations are the reactor protection system, reactor kinetics and reactivity feedback coefficients. Analyses presented in this category are the Turkey Point Uncontrolled RCCA Withdrawal transient benchmarked to FSAR results (Section 6.1), and the St. Lucie Unit 2 CEA Drop transient benchmarked to FSAR results (Section 6.2).

6.1 Turkey Point Uncontrolled RCCA Withdrawal 6.1.1 Transient Descri tion A slow, uncontrolled,rod cluster control assembly (RCCA) withdrawal transient from 100% power was simulated with the RETRAN02 computer code and benchmarked to the analogous transient documented in the Turkey Point FSAR.. (Ref. 13). In this transient the rod withdrawal causes an increase in core power and heat flux which result in increases in RCS temperature and pressure. Reactor trip can occur on high RCS pressure, high pressurizer level or on exceeding the high power, overpower hT or overtemperature QT setpoints. This transient assesses, the adequacy of the RETRAN reactor kinetics modeling and the modeling of the reactor protection system.

ATTACHMENT 2 Page 2 of 8 6.1.2 RETRAN Anal sis Descri tion The initial conditions of the benchmark and RETRAN02 analysis, presented in Table 6.1.1, were incorporated into the Turkey Point RETRAN base model (see Appendix B). These initial conditions represent beginning of cycle conditions as listed in the Turkey Point FSAR. The analysis was performed for a rod withdrawel rate;of 2e5xl0-5~%/sec. For this case the reactor trips on overtemperature hT. Presented in Table 6.1.2 is the status of safety systems included in the RETRAN simulation of this transient.

6.l.3 Results Results of the RETRAN02 calculation and the FSAR are presented in Figures 6.1.1, 6.1.2 and 6.1.3. A sequence of events for both the RETRAN calculation and the FSAR results is V

shown in Table 6.1.3. As the core power increases the sensed temperature difference between the hot leg and cold leg reaches the dynamic overtemperature 5T setpoint, when the scram signal is generated and the reactor trips. The turbine trips on the reactor trip. RETRAN predicts the reactor trip about 3 seconds later than the FSAR analysis but maximum core power, maximum RCS pressure and temperature calculated for the RETRAN and FSAR analyses are essentially identical. Overall, the RETRAN simulation shows good agreement with the FSAR analysis results.

ATTACHMENT 2 Page 3 of 8 TABLE 6.1.1 INITIAL CONDITIONS AND KEY PAEVQ4ETERS UNCONTROLLED RCCA WITHDRAWAL PAEVQiETER Core Power, MW (Thermal) 2244 Core Inlet Coolant Temperature, 'F 550.2 Core Mass Flow Hate, 106 ibm/hr 101.5 Pressurizer Pressure, psia 2220 Doppler Coefficient, 10 45.k /'F .12 Moderator Tem'perature Coef f icient, 10 4 5 8/'F ~ 4 Over-Temperature 4T Above Nominal dT Trip setpoint (%)

Red witMrawal Rate a.0/<<c- 2.5X10

lg ATTACHMENT 2 Page 4 of 8 TABLE 6.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL UNCONTROLLED RCCA WITHDRAWAL AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM) X Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System X S. G. Level Control System Automatic Rod Motion

ATTACHMENT 2 Page 5 of 8 TABLE 6. l. 3 SEQUENCE OP EVENTS UNCONTROLLED RCCA WITHDRAWAL TIME(S ) PARAMETER F SAR RETRAN F SAR RET RAN Rod Withdrawn 0 0 Reactor Tripped on 50.5 55.6 Over-Temp. QT Turbine Tripped on 56.6 Reactor Trip Maximum Core Power 51 55 ll3.6% 113.4%

Maximum Pressurizer 51 55 2332 psia 2322 psia Pressure Maximum Core Average 52 57 585.8 F 585.6 F Temperature

cn C4 M O cn lU C4 RETRAN 0 FSAR O

tA ea IO 20 30 40 50 60 7 I NE (SE C)

FIGURE 6.1.1 PRESSURIZER PRESSURE UNCONTROLLED RCCA WITHDRAWAL

RETRAN 0 FSAR 20 30 T I NE (SEC)

FIGURE 6,1.2 PERCENT CORE POWER UNCONTROLLED RCCA WITHDRAWAL

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A r f~

q

RHTRAN FSAR 10 ~0 30 50

~ I NF (SEC)

FIGURE 6.1.3 AVERAGE CORE COOLANT TEMPERATURE UNCONTROLLED RCCA WITHDRAWAL

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