ML17252A753

Lpci/Containment Cooling Sys Evaluation.

ML17252A753
Person / Time
Site:	Dresden
Issue date:	11/30/1992
From:	Mintz S, Torbeck J GENERAL ELECTRIC CO.
To:
Shared Package
ML17179A771	List: ML17179A770 ML17179A772 ML17252A753 ML20012G662
References
DRF-T23-685, GENE-770-26-109, GENE-770-26-1092, NUDOCS 9303110228
Download: ML17252A753 (116)

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GENE-770-26-1092 DRF-T23-685 CLASS II November 1992 --*-

Dresden Nuclear Power Station Units 2* and 3 LPCI/Containment Cooling System Evaluation Prepared by:

S. Hi ntZ P7ant Performance Analysis Projects Approved b~~~ . E. Torbeck Plant Performance Analysis Projects

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GENE-770-26-1092 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT The only undertakings of the General Electric Company (GE) respecting information in this document are contai.ned in the contract between Conunonwealth Edison Company (CECO) and GE, as identified in Purchase Order Number 341715 YY25, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than CECO, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liabili.ty as to the completeness, accuracy, or, usefulness of the information contained in this document~ or that its use may not infringe privately owned rights.

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GENE-770-26-1092 ABSTRACT This report provides the results of an evaluation of the Dresden containment response during a design basis loss-of-coolant accident (DBA-LOCA) considering the current Dresden LPCl/Containment Cooling System parameters. The results of the Dresden containment pressure and temperature response analysis described in this report can be used to upda~e the Dresden SAR and thus clarify the SAR assumption on the number of Containment Cooling Service Water (CCSW) pumps for the li~iting containment cooling case in SAR Section 5.2.

This report alio contains a revie~ of a NFS Cal~ulation RSA-D-92-01 which was provided to GE by CECO to determine the impact of the suppression pool "temperature results documented in NFS Calculation RSA-D-92-01 on the temperature data used in evaluating the containment dynamic loads defined during the Mark I Long Term Program on torus attached piping.

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GENE-770-26-1092 TABLE OF CONTENTS

• ABSTRACT

1.0 INTRODUCTION

1 2.0 CONTAINMENT PRESSURE ANO TEMPERATURE RESPONSE 4 2.1 Model Description 4 2.2 Analysis Assumptions 4 2.3 Analysis Description 5 2.4 Results 6

3.0 CONCLUSION

S 8

4.0 REFERENCES

9 APPENDICES A. REVIEW OF NFS CALC RSA-D-92-01 A-1 B. REDUCTION TO THE CONTAINMENT PRESSURE 8-1 C. PRIMARY SYSTEM MASS AND ENERGY RELEASE C-1 DATA FOR DRESDEN CONTAINMENT EVALUATION

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GENE-770-26-1092

• 1.0 Introduction Section 5.2 of the Dresden SAR documents long-term heatup analyses performed to evaluate the capability of the Dresden LPCI/Containment Cooling System to maintain peak containment pressures and temperatures within limits during the design basis loss-of-coolant accident {DBA-LOCA}. The DBA-LOCA for the Dresden Plant is a dou?le-ended guillotine break of a recirculation suction line. Four cases for different LPCl/Containment Cooling configurations are described in Section 5.2 of the SAR. Table 1 summarizes the LPCI/Containment Cooling

• **parameters: for . these .four cases in the SAR. It was recently determined that the measured Containment Cooling Service Water {CCSW} flow rate during two pump operation for a single LPCl/Containment Cooling System Loop is less. than the value used in the SAR analysis. This would result in a decrease in the

• • LPCl/Containment Cooling System heat exchanger performance and therefore result in higher peak suppression pool temperatures. To assess the impact of reduced heat exchanger performance, long-term analysis of the containment pressure and

• temperature after initiation of the LPCl/Containment Cooling System {600 seconds into the event} was performed. Since the SAR reports that 2 CCSW pumps per heat exchanger are assumed in the SAR analysis for all 4 cases, the limiting case for one loop with.two CCSW pumps in operation was re-analyzed with the reduced CCSW flow rate. Both Case 1 and Case 3 have this configuration, and the SAR reports the same peak temperature {see SAR Figure 5.2.3:3) for both cases. Case 3, which assumes only 1 Core Spray pump is available, was chosen for the re-analysis. Case 4 of Section 5.2 of the SAR which produced the maximum temperature of the four SAR cases was also described as using 2 CCSW pumps. However, a review of the Dresden SAR and GE files indicated that the analysis used to produce the response for Case 4 in Section 5.2 of the SAR assumed only 1 CCSW pump. Therefore, Case 4 was reanalyzed for this report with the assumption that only 1 CCSW pump is GENE-770-26-1092 available. The analyses which are documented in this report use the current values of the CCSW and LPCI/Containment Cooling flow rates through the heat exchanger and the current heat exchanger performance, which are described in References 1,2 &3 {with and without flow rate reductions to account for uncertainties in the flow rates). The containment pressure and temperature response analysis described in this report was performed in accordance with Regulatory Guide 1.49 using current GE codes and models {References 4,5 &6).

In addition to the evaluation of the Dresden LPCI/Containment Cooling System

.described above, Appendix A to this report contains a review of NFS Cale.

RSA-D-92-01. The purpose of the review was to determine the impact of the results of NFS Cale. RSA-D-92-01 on the temperature data used in evaluating the containment dynamic loads defined for torus attached piping in Dresden during the Mark I Long Term Program {LTP).

Appendix B gives an estimate of the reduction to the containment pressure at the time of the peak suppression pool temperature, for initial conditions which minimize the containment pressure response. A request for this information was

• • made by the Commonwealth Edison Company (CECO) in discussions with General Electric (GE) during the course of the program to evaluate the LPCI/Containment Cooling System.

Appendix C provides the mass and energy release data obtained from the Dresden

. containment analysis described in Section 2.0.

Resylts summary The peak suppression pool temperature for SAR Case 3 {2 LPCI/Containment Cooling**System pumps and 2 CCSW pumps) is 3°F higher than the SAR value of 165°F when the uncertainty in the LPCl/Containment Cooling System and CCSW flow rates is not accounted for and 6°F higher than the SAR value when the uncertainty in the flow rates is accounted for. The peak suppression pool temperature for SAR Case 4 (1 LPCl/Containment Cooling System pump and 1 CCSW pump) is equal to the SAR value of 180°F when the uncertainty in the flow rates GENE-770-26-1092 is not accounted for and 6°F higher than the SAR value when the uncertainty in the flow rates is accounted for.

The review of NFS Cale. RSA-D-92-01 confirmed that the results of the NFS calculation do not impact the temperature data used to evaluate the Mark I containment loads ~pecified for torus attached piping during the Mark I LTP; GENE-770-26-1092

2. O Containment Pressure and Temp*erature Response 2~1 Model Description*

A coupled reactor ~ressure vessel and containment model, based on the Reference 4 and Reference 5 models, was used to calculate the long-term (> 600 seconds) transient response of the containment during the DBA-LOCA. This model performs fluid mass and energy balances on the reactor primary system and the suppression pool, and calculates the reactor vessel water level, the reactor vessel pressure, the pressure and temperature in the drywell and suppression chamber airspace and the bulk suppressi~n pool temperature. The various modes of operation of all important auxiliary systems, such as SRV's, the MSIV's, ECCS, the RHR system (LPCI/Containment Cooling system in the case of Dresden) and feedwater are modeled. The model can simulate actions based on system *

setpoints~ automatic actions and operator-initiated actions.

2.2 Analysis Assumptions The initial conditions and key input parameters used in the analysis are provided in-Table 2. These are based on the current Dresden containment data which are documented in References 2 &7. The following key input assumptions were used in performing the Dresden containment LOCA pressure and temperature response analysis:

I.* The reactor is operating at 102% of the rated thermal power.

2. Vessel blowdown flowrates are based on the Homogeneous Equilibrium Model

{Reference 6).

3. The core decay heat is based on ANSl/ANS-5.1-1979 decay heat (Reference

8)

• GENE-770-26-1092

4. Feedwater flow into the RPV continues until all the feedwater above ISO"F is injected into the vessel.

5. Thermodynamic equilibrium exists be~ween the liquids and gases in the drywell. Mechanistic heat and mass transfer between the suppression pool and the suppression chamber airspace is assumed.

6. The vent system flow to the suppression pool consists of a homogeneous mixture of the fluid in the drywell.

7. The initial suppression pool volume is at the minimum Technical Specification (T/S) limit to maximize the calculated suppression pool temperatu.re .

. 8. The initial suppression pool temperature is at the maximum T/S value to maximize the calculated suppression pool temperature.

9. Consistent with the SAR, containment sprays are used to cool the containment.

10. Passive heat sinks in the drywell, suppression chamber airspace and suppression pool are conservatively neglected.

• 11. All Core Spray and LPCl/Containment Cooling System pumps have 1003 of

their horsepower rating converted to a pump heat input which is added either to the RPV liquid or suppression pool water.

12. Heat transfer from the primary containment to the reactor building is co1iservatively neglected.

2.3 Analysis Description The long-term containment pressure and temperature response was analyzed GENE-770-26-1092 for the DBA-LOCA which was identified in the SAR as an instantaneous double-ended guillQtine break of a recirculation suction line. Case 3 and Case 4 of Section 5.2 of the SAR (Curves C and D in SAR Figure 5.2.3:2 and Figure 5.2.3:3) were re-analyzed .for this report. Case 3 in Section 5.2 of the SAR is used to establish the long-term design basis pool cooling temperature conditions. The LPCI/Containment Cooling System parameters for Case 3 are consistent with the Auxiliary Systems Data Book (Reference 9) and Mode Bon the process diagram (Reference 10). For Case 3 it is assumed that one loop, with one heat exchanger, two LPCI/Containment Cooling System pumps and two CCSW pumps are available. Case 4 as described in the SAR assumes the availability of one LPCI/Containment Cooling System pump and two CCSW pumps. For the analysis of thjs report it was assumed that only 1 CCSW pump is available.

This is consistent with the number of CCSW pumps reported for Mode C in the Process Diagram. Additional analyses (identified as Cases 3A and 4A in this report) were performed with a lower heat exchanger heat removal rate to account for the uncertainty in the LPCI and CCSW flow measurements. *Table 3 sunvnarizes the LPCI/Containment Cooling System parameters assumed for the long-term heatup analyses of this report (References 2,7). Appendix C provides break flow mass and energy data for the analysis. Note that the integrated break flow mass and energy given in Appendix C is an output from the coupled vessel and containment model used for the analysis.

2.4 Results Table 4 summarizes the results of the long-term heatup calculations. Figures 1, IA, 18, 2, 2A and 28 show long-term pressure and temperature response for Cases 3 and 4, respectively, with the assumption of nominal flow rates.

Figures 3, 3A, 38, 4, 4A and 48 show the containment pressure and temperature responses for Cases 3 and 4 obtained with the reduced heat exchanger K values which account for flow measurement uncertainty. The results in Table 4 show

• that the peak pool temperature with the nominal flow rates for Case 3 is 3*F higher than the SAR value shown in Figure 5.2.3:3 of the SAR while the peak suppression pool temperature for Case 4 is unchanged. This difference in the results between Case 3 and Case 4 is attributed to the reduction in the CCSW GENE-770-26-1092 flow rate for 2 pump operation in Case 3 to 5600 gpm versus the SAR value for

• Case 3 of 7000 gpm. Note that SAR Figure 5.2.3:3 shows the drywell temperature

. only. However, during.the time of the peak suppression pool temperature, the drywell and suppression pool temperature w~ll be nearly the same.* The difference in the peak pool temperature between Case 3A and Case 4A of *this report is the same as the difference between Cases 3 and 4 in the SAR. This indicates that only 1 CCSW pump was originally used for Case 4 of the SAR.

There is a significant effect on the peak suppression pool temperature of using a heat exchanger K value which accounts for uncertainty in the LPCI/Containment Cooling and CCSW flow rates. The increases in the peak suppression pool temperatures due to the use of the reduced*K values are 3°F for Case 3,3A and 6°F for Case ~,4A .

GENE-770-26-1092 3.0 Conclusions

• The peak suppression pool temperatures based on the use of nominal values of the current LPCl/Containment Cooling and CCSW flow rates through the LPCl/Containment Cooling System heat exchanger result in peak suppression pool temperatures which are 0 to 3°F higher than the SAR values. The use of decreased heat exchanger coefficient values to account for the uncertainty in the LPCl/Containment Cooling and CCSW flow rates result in peak suppression pool temperatures which are 6°F higher than the results with the nominal pump flow rates and which are also 6°F higher than the values reported in Section 5.2 of the Dresden SAR .

GENE-770-26-1092 4.0 References

• 1) Letter, G. G. Chen to S. Mintz,"K Values for Dresden Units 2 & 3 Containment Heat Exchangers," September 14, 1992.

2) Letter, C; R. Parker to S. Eldridge (CECO),"LOCA Long-Term Containment Response Analysis K-values for LPCl/Containment Cooling System Heat Exchangers Dresden Nuclear Power Station, Units 2 &3," October 6, 1992.

3) Letter, S. L. Eldridge/B. M. Viehl to T. Allen, "Inputs for Heat Exthanger Parameters for CCSW Flow Issue Dresden Units 2 &3," August 31, 1992.

4) NEDM-10320,"The GE Pressure Suppression Containment System Analytical

• Model," March 1971.

5) NED0-20533,"The General Electric Mark III Pressure Suppression Containment System Analytical Model," June*l974.

6) NED0-21052,"Maximum Discharge of Liquid-Vapor Mixtures from Vessels,"

General Electric Company, September 1975 .

. 7) Letter, C.R. Parker to S. Eldridge (CECO),"LOCA Long-Term Containment Response Analysis Input Parameters Dresden Nuclear Power Station, Units 2

&3 (Final Values)," September 21, 1992.

8) "Decay Heat Power in Light Water Reactors," ANSI/ANS 5.1 - 1979, Approved by American National Standards Institute, August 29, 1979.

9) Auxiliary Systems Data Book, Plant Dresden 2, GE Document 257HA654, Issued April 15, 1979.

10) LPCI Containment Cooling System Process Diagram, GE DWG 729E583, Rev. 1, February 24, 1969.

GENE-770-26-1092 Table 1 - Flow Rates Used in SAR Containment Response Analysis

• Case No. of Loops**

LPq/

Containment Cooling*

Pumps Per Loop Total LPCI/

Containment Cooling Pump Flow (qpml ccsw Pumps Total ccsw Pump Per Loop Flow (gpml 1 1 2 10,000 2 7000 2 2 2 20,000 2 14000 3 1 2 lO, 000 2 7000 4 1 1 5,000 2* 7000*

• Section 5.2 ~f the SAR reports that two CCSW pumps/HX were assumed for Cases 1 to 4. However, it is believed that only one CCSW pump was used for the original analysis for SAR Case 4.

'** 1 Heat Exchanger (HX) per LPCI/Containment Cooling Loop.

GENE-770-26-1092 Table 2 - Input Parameters Used for Containment Analysis Value Used in Parameter Units Analysis Core Thermal Power. MWt 2578 Vessel Dome Pressure psi a 1020 Drywell Free (Airspace) Volume ft 3 158236 (including vent system)

Initial Suppression Chamber Free (Airspace) Volume Low Water Level (LWL) ft 3 120097 Initial Suppression Pool Volume

,Min. Water Level ft 3 112000 Initial Drywell Pressure psig 1.25 Initial Drywell Temperature *F 135 Initial Drywell Relative Humidity  % 20 Initial Suppression Chamber Pressure psig 0.15 Initial Suppression Chamber Airspace Temperature *F 95 Initial Suppression Chamber Airspace  % 100 Relative Humidity

.Initial Suppression Pool Temperature 95 No. of Oowncomers 96 Total Downcomer Flow Area ft 2 301.6 In it i a1 *Oowncomer Submergence ( LWL) ft 3.67 GENE-770-26-1092 Table 2 - Input Parameters Used for Containment Analysis

• Parameter Downcomer I.D.

Vent System Flow Path Loss Coefficient ft Value Used in Analysis 2.00 (includes exit loss) 5.17 Supp. Chamber (Torus) Major Radius ft 54.50 Supp. Chamber (Torus) Minor Radius ft 15.00 Suppression Pool Surface Area ft 2 9971.4 (in contact with suppression chamber airspace)

Suppression Chamber-to-Drywell Vacuum Breaker Opening Diff. Press.

- start psid 0.15

- full open psid 0.5 Supp. Chamber-to-Drywell Vacuum Breaker Valve Opening Time sec 1.0

. Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve ft 2 3 .14 assembly)

Supp. Chamber-to-Drywell Vacuum Breaker Flow Loss Coefficient *

(including exit loss) 3.47 No. of Supp. Chamber-to-Orywell Vacuum Breaker Valve Assemblies (2 valves per assembly) 6 LPCI/Containment Cooling Heat Exchanger Kin Containment Cooling Mode Btu/s-°F See Table 3 LPCI/Containment Cooling Service Water Temperature GENE-770-26~1092 Table 2 - Input Parameters Used for Containment Anaiysis Value Used in Parameter Analvsis LPCl/Containment Cooling Pump Heat (per pump) hp 700 Core Spray Pump Heat (per pump) hp 800 Time for Operator to turn on LPCl/Containment Cooling System in Containment Cooling mode (after LOCA signal) sec 600 Feedwater Addition (to RPV after start of event; mass and energy)

Feedwater Mass Enthalpy

• Node ** 1l!2ml CBtu/lbml 1 34658 308.0 2 96419 289.2 3 145651 268.7 4 91600 219.8 5 *65072 188.4

• Includes sensible heat in the feedwater system pipe metal.

• • Feedwater mass and energy data combined to fit into 5 nodes for use in the analysis.

GENE-770-26-1092 Table 3 - LPCl/Containment Cooling System Parameters Used in Analysis Total LPCI/ LPCI/

Containment Containment Total Cooling Cooling No. of ccsw HX No. of

• Pumps Flow ccsw Pump K

. Case Loops* Per Loop (qpml Pumps Flow (gpml CBtu/s-*Fl 3 1 2 10,000 2 ,5' 600 356.1 3A** 1 2 8,916 2 4,795 327.3 4 1 1 5,000 1 3,500 249.6 4A** 1 1 3,881 1 3,071 219.2

• one-.heat exchanger per loop

• • with the uncertainty in the LPCI/Containment Cooling and CCSW flow rates accounted for*

GENE-770-26-1092 Table 4 - Peak Suppression Pool Temperature~

• Case No.

Maximum Suppression Pool Temperature (*Fl FSAR Temperature ( *F)*

3 168 165 3A 171 N/A 4 180 180 4A 186 N/A

• Note that the FSAR reported drywell temperatures and not suppression pool

'*temperatures ... However, during the times of peak suppression pool temperature

'the drywell and pool temperatures should be similar.

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. 2. 3-13 Again the core heatup and extent of metal-water reaction are as discussed. The containment pressure and temperature are shown as curve "d" in Figures 5.2.3

2 and 5.2.3:3 respectively. It is shown that following the initiation of the single containment spray cooling pump and its associated

- heat exchanger, the* containment pressure decreases initially, then slowly increases to the maximum shown in Table 5.2.3:1 due to addition of decay-energy to the containment. Thereafter, energy removal by the single containment spray cooling pump and heat exchanger exceeds the addition rate from all so.urces, resulting in decreasing containment pressure.

Containment spray itself does not significantly affect the peak post accident pressure rise. It does result in a somewhat faster depressurization i11111ediately following the completion of the blowdown, however. The controlling parameter affecting the post accident secondary peak in pressure is the heat removal capability of the containment cooling

rn~v+ i heat exchanger relative to the core decay heat production.

In~rf 2.. 5.2.3.4 Containment Capability with Respect to Metal-Water Reactions

( s~ e.1-7 C>Yl s.i.-:;.3.1) A. Nature of Requirements If zircaloy of the reactor core is heated above about 2ooo*r in

.the presence of steam due to an accidental loss-of-coolant, a chemical reaction occurs in which zirconium oxide and hydrogen are formed. This is accompanied by an energy release of about 2800 BTU per pound of zirconium reacted. The energy produced is accomodated in the suppression chamber pool. The hydrogen formed, however, w_ill result in an increased pressure due simply to the added moles of gas in the fixed volume depending on the amount produced. Although very small quantities of hydrogen are produced with core spray, the containment has the inherent ability to accommodate much larger amounts as discussed below.

B. Expected Metal Water Reactions The metal-water reactions during core heatup, and within the first 40 to 60 minutes during which portions of the core are at temperatures of interest in metal-water reactions, are calculated

_by a core heat-up computer code. The core is sub-divided into nodes consisting of 5 radial zones, five axial nodes, 4 relative rod powers within each assembly, and with 4 radial fuel nodes in each fuel rod. Heat-up is calculated during the blowdown phase employing experimental heat transfer coefficients. Unde~ core spray conditions experimentally determined coefficients from prototypetests are applied. The metal water reaction is calcylated as each node temperature is determined by the parabolic law. This is integrated over the entire core until the rods are finally wetted and cooled by the core spray system about an hour after the accident. The extent of the metal-water reaction thus calculated is oxidation of under 0.51 of all the zirconium in the core. This reaction produces an additional energy release of only

• 1ANL6548, "Studies of Metal-Water Reaction at High Temperatures III Experimental and Theoretical Studies of Zirconium-Water Reaction."

ZDFSAR/32

• The containment pressure and temperature response after initiation of containment sprays for curves c and d have been recalculated using updated L..PCI/Containment Cooli-ng System parameters. The description of this updated analysis is given in Section 5.2.3.3.1 .

5.2.3.3.1 Updated Contairiment Characteristics After Reactor Slowdown Flow measurements 1 have determined that the measured Containment Cooling

• S$rvice Water (C~SW) flow rate during two pump operation for a single LPCl/Containment Cooling System Loop is less than the value assumed in the analysis which produced the pressure and temperature curves in Figures 5.2.3:2 and 5.2.3:3. This *would result in a decrease in the LPCl/Containment Cooling System heat ex~hanger performance and, therefore, result in higher peak containment temperatures. Therefore the impact of reduced heat exchanger performance was assessed with long-term analyses of the containment pressure and temperature response after initiation of the LPCl/Containment Cooling System (600 seconds into the event). The limiting case with two CCSW pump operation, Case 3, was re-analyzed with the reduced CCSW flow rate. Case 4, which produced.the maximum temperature, was also originally described as using 2 CCSW pumps. However, a review of vendor files3 indicated that the analysis

.used to produce the response for Case 4 assumed only 1 CCSW pump. Therefore, Case 4 was reanalyzed with t~e assumption that only 1 CCSW pump is 4Vailable.

The analysis for Case 3 and Case 4 used values of the CCSW and LPCI/Containment Cooling flow rates through the LPCl/Containment Cooling heat exchangerl and

• values of the heat exchanger performance which accounted for the uncertainty in the LPCI/Containmnent Cooling and CCSW flow ratesz.

A coupled reactor pressure vessel and containment model, based on the General Electric containment models4.s, was used to calculate the transient response of the containment during the DBA-LOCA. This model performs fluid mass and energy balances on the reactor primary system, the drywell airspace, the suppression chamber airspace and the suppression pool, and calculates the reactor vessel water level, the reactor vessel pressure, the pressure and temperature in the drywell and suppression chamber airspace and the bulk suppression pool temperature. The various modes of operation of all important auxiliary systems, such as SRV's, MSIV's, ECCS, LPCI/Containment Cooling System and feedwater are modeled. The model can simulate actions based on system setpoints, automatic actions and operator-initiated actions

• The initial conditions and key input parameters used in the analysis are provided in Table 5.2.3:2. Table 5.2.3:3 summarizes the LPCI/Containment

• Ceu>ling System paramet~rs assumed for the long-term heatup analysis. The following key input assumptions were used in performing the analysis:

1. The reactor is operating at 102% of the rated core thermal power.

2. Vessel bl~wdown flow rates are based on the Homogeneous Equilibrium Model 6 *

3. The core decay heat is based on ANSI/ANS-5.1-1979 decay heat 7 *

4. Feedwater*flow into the RPV continues until all the feedwater above 180.F is injected into the vessel.

5. Thermodynamic equilibrium exists between the liquids and gases, in the drywell. Mechanistic heat and mass transfer between the suppression pool and the suppression chamber airspace is assumed. *

6. The vent system flow consists of a homogeneous mixture of the fluid in the drywell.

7. The initial suppression pool volume is at the minimum* Technical Specification value to maximize the calculated suppression pool temperature.

8. The initial supp~ession pool temperature is at the maximum Technical Specification value to maximize the calculated suppression pool tem~erature.

9. Containment sprays are used to cool the containment .

10. Passive heat sinks in the drywell, suppression chamber airspace and suppression pool are conservatively neglected.

11. All Core Spray and LPCl/Containment Cooling System pumps have 100% of their horsepower rating converted to a pump heat input which is added either to the*RPV liquid or suppression pool water.

12. Heat transfer from the primary containment to the reactor building is conservatively neglected.

Results Table 5.2.3:4 summarizes the results of the long-term heatup calculations.

Figures 5.2.3:7 to 5.2.3:9 show long-term pressure and temperature response for Case 3 and Figures 5.2.3:10 to 5.2.3:12 show the pressure and temperature

• response for Case 4. The results in Table 5.2.3:4 show that the peak suppression pool temperatures are higher than the values shown in Figure 5.2.3:3 of the _SAR. Note that SAR Figure 5.2.3:3 shows the drywell temperature only. However, during the time of the peak suppression pool temperature, the drywell and suppression pool temperature will be nearly the same. The difference in the peak pool temperature between Case 3 and Case 4 in Table 5.2.3:4 is the same as the difference between Curves c and d in Figure 5.2.3:3.

This confirms that only 1 CCSW pump was originally used to determine the containment pressure and temperature response for Case 4 (Curved) .

References:

1) Letter, S. L. Eldridge/8. M. Viehl (CECO) to T. Allen (GE), "Inputs for Heat Exchanger Parameters for CCSW Flow Issue Dresden Units 2 &3," August 31, 1992.

2) GE Report GENE-770-26-1092,"D~esden Nuclear Power Station - Units 2 and 3, LPCI Containment cooling System Evaluation," November 1992.

3) Letter, S. Mintz (GE) to J. E. Nash (GE),"Design Basis for LPCl/Containment Cooling System Heat Exchanger Sizing," April 6, 1992.

4) NEDM-10320,"The GE Pressure Suppression Containment System Analytical

~del," G*eneral Electric Company, March 1971.

5) NED0-20533,"The General .Electric Mark III Pressure Suppression Containment System Analytical Model~" General Electric Company, June 1974~

6) NE00-21052,"Maximum Discharge of Liquid-Vapor Mixtures from Vessels,"

• Genera 1 Electric Company, September 1975.

7) "Decay Heat Power in Light Water Reactors," ANSl/ANS-5.1 - 1979, Approved by American National Standards Institute, August 29, 1979 .

Table 5.2.3:2 - Input Parameters Used for Containment Analysis Value Used in Parameter Analysis Core Thermal Power MWt 2578 Vessel Dome Pressure psi a 1020 Drywell Free (Airspace) Volume ft3 158236 (including vent system)

Initial Suppression Chamber Free (Airspace) Volume Low Water Level (LWL) ft3 120097 Initial Suppression Pool Volume Min. Water Level ft3 112000 Initial Orywel l Pressure psig 1.25 Initial Drywel l Temperature *F 135

. Initial Drywell Relative Humidity  % 20 Initial Suppression Chamber Pressure psig 0.15 Initial Suppression Chamber Airspace Temperature *f 95 Initial Suppression Chamber Airspace  % 100 Relative Humidity Initial Suppression Pool Temperature . *f 95 No. of Downcomers 96 Total Downcomer Flow Area ft 2 301.6

• Initial Downcomer Submergence (LWL) ft 3.67

Table 5.2;3:2 - Input Parameters Used for Containment Analysi.s Value Used in Parameter Analysis Downcomer I . D. - ft 2.00

• vent System Flow Path Loss Coefficient (includes exit loss) 5.17 Supp. Chamber (Torus) Major Radius ft 54.50 Supp. Chamber (Torus) Minor Radius ft 15.00 Suppression Pool Surface Area ft 2 9971.4 (in contact with suppression chamber airspace)

Suppression Chamber-to-Drywell Vacuum Breaker Opening Diff. Press.

- start psid 0.15

• * -

• fu 11

• open psid 0.5 Supp. Chamber-to-Drywell Vacuum Breaker Valve Openini Time sec 1.0 Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve ft 2 3.14 assembly)

Supp. Chamber-to-Drywell Vacuum Breaker Flow Loss Coefficient (including exit loss) 3.47 No. of Supp. Chamber-to-Drywell Vacuum Breaker Valve Assemblies

• * (2 va*l ves *per assembly) 6 LPCl/Containment Cooling Heat Exchanger K in Containment Cooling Mode Btu/s-*F See Table 5.2.3:3 LPCl/Coritainment Cooling Service Water Temperature

Table 5.2.3:2 - Input Parameters Used for Containment Analysis Value Used in Parameter Analysis LPCI/Containment Cooling Pump Heat (per pump) hp 700 Core Spray Pump Heat (per pump) hp 800 Time for Operator to turn on LPCl/Containment Cooling System in Containment Cooling mode (after LOCA signal) sec 600 Feedwater Addition (to RPV after start of event; mass and energy)

Feedwater Mass Enthalpy

• Node ** 1JJ2ml (Btu/lbml 1 34658 308.0 2 96419 289.2 3 145651 268.7 4 91600 219.8 5 65072 188.4

• Includes sensible heat in the feedwater system pipe metal.

• • Feedwater mass and energy data combined to fit into 5 nodes for use in the analysis.

Table 5.2.3:3 - LPCI/Containment Cooling System Parameters Used in Analysis of Section 5.2.3.3.1

• Case No.

No. of Loops*

LPq/

Containment Cooling Pumps Per Loop LPCI/

Containment Cooling No. of Flow**

(gpml ccsw Total ccsw Pump Pumps Flow (qpml HX K

(Btu/s-*F) 3 1 2 8,916 2 4,795 327.3 4 1 1 3,881 1 3,071 219.2

• There is one heat exchanger per loop.*

• • This is *the LPCl/Containment Cooling System flow rate after 600 seconds and it is used in the containment spray mode.

Table 5.2.3:4 - Peak Suppression Pool Temperatures With Updated Containment Cooling Parameters Peak Suppression Pool Case No. Temperature ("Fl 3 171 4 186

• Table 5.2.3:5 Available NPSH for LPCI Pumps Post DBA LOCA Case Total Single Torus Torus Static Specific Vapor Suction NP SHA NPSHR Flow Pump Temp Pressure Head Volume Pressure Piping (ft) (ft) 3

. (gpm) Flow (oF) (psia) (ft) (ft /lb) (psi a) Losses (gpm) (ft) 3 8,916 4,458 171 19.1 13.32 0.016457 6.1318 3.15 40.3 26.9 4 3,881 3,881 186 20.6 13.32 0.016547 8.568 2.27 39.72 25.7

Reference:

Calculation NED-M-MSD-43

I I

60. --*-- *--*

I l 1 Orywell 2 Suppression Cha111L er

'10.

a:

t--t (f)

Q_

20.

- L ~ 2 _ __L L _

w ...

a: ...

l (f) -- *

(/)

w -

a: -

Q_ -

0. I I I I I I I I I 10 102 101 104 figure~

~.l.

- Long- ler* OBA-LOCA Orywel J. an 1; 7 TIME - SEC preuton Chiltlber Pressure Ae>pon>e tor ("'* J/.

TEMPER~TURE DEG F

-8 8 8

.=

.8 0 1 I I I I I I I

-~ --

~

c

~

!-f\ ! ~

l.,J *.

./ft w~*

Ge'

~

~

I I

0 N

-4 I

!,.. I

,..~ .;.....-/

1 c

I

(

I I-

~

I c

ID I I

I

~

fD c

'e "'c

I

"*c"'-. .

.. .. I I

I"">

I i

.w "  !

i I

I i

I I

! I

i .

t~t*t*s re 3H'J .JOJ asuodsaff a..1n1uadwa1 (00d UOf sn..1ddns YJ01-V80 ...... :-6uo1 -_,,< 8..lhfitJ lJS - 11111 zOI 01

' ,. ' ' I ' ' ' '- *o

- -f

- fTI

-- *~

,. rn

-u

- :JJ

-- D

-t

• 001 c

D rn

--- - -*- *---.-------~.

\

60. --- .. -**

I I

1 Orywe l l 2 Suppression Cha ml ,.!f

'10.

~

(I

( fl Q_

\ ~..._. .2

• l w

20.

a: ...

J ...

(fl ...

(/)

w a:.

Q_ -

0. l l l l I I l l I 4

10 101 10 TIME - SEC figure.A' - Lon9-len1 OBA-LOCA Orywell *nd Suppre55ton CbUlber Pressure Response for C~~e 4.(

E;. 1 . 3. I 0

400.

! I I

300.

LL el w

Cl w

a:

> 200.

l-a:

\ ...

I

• cc ....

w ....

Q_ ~

~ ._

w ._

t- ._

100. I I I I I I I I I 10 102 104 10 5

) .

THIE - SEC figure~ - long-lena DBA-LOCA Drywell Te11periture Response for Case 41

"'5,;i.1.11

300. *-*-* .

l 200.

LL e> . ~.

w D

.,,.,.- ~

w a:

l 100. -**- ---- .. ..

t- ....

a: ....

a: ...

w ...

Q_

~

w I-

0. I I I I I I I l l '

10 to2 10 3

10 4

TIME - SEC figure~ - long-Tena OBA-LOCA Suppress ion Pool Temperilure Response for. Ca~e ~

. ~*s-.-:i.1:12 .

6.2.4-4

• TABLE 6.2.4:1 LPCI/CONTAINMENT COOLING EQUIPMENT SPECIFICATIONS Main System Pumps Number 4 (3 required to meet design basis)

Type Single stage, vertical, centrifugal Seals Mechanical Drive Electric rnotor Power source Nonnal auxiliary or emergency diesel Speed 3600 rpm Pump casing Cast steel Impeller Bronze Shaft Stainless steel Code ASME Section III B Perfonnance Characteristics - 3 pumps running At 0 psi reactor pressure Flow 5350 gpm each - 16,000 gpm total Head 263 feet

.-JiP~ow~etr~~~~E~~~~~~go~o~~h~p~each

~H !@variable) ii.~

- 1800 hp total At 200 psi. reactor pressure Flow 2675 gpm each - 8,000 gpm total Head 565 ft 1500 hp to ta 1 dBZr (twai 1able) 490 ~each -

4"1 Perfonnance Characteristics - 1 pump running f

Flow

• 5990 gpm Head 135 ft Power 560 hp c]P~ (Available) 40 ft).

Containment Cooling Service Water Pumps Number 4 ( 2 needed to provide required cooling capacity)

Type Horizontal, centrifugal Po:>.,1er source Auxiliary transfonner or emergency diesel Capacity 3500 gpm each - 7000 gpm total Head (approximately) 435 ft

\co(.¥\' ~

June 1992 6.2.4-5 TABL! 6.2.4:1 LlCI/CONTAINMl1fT CQOLUfG EQUIPMENT SPECI~ICAIIQMS (Contd.)

Containment Cpolin1 Scryicc Water Pymp1 Number 4 (2 needed to provide required cooling capacity)

Type Horizontal, centrifugal Power source Auxiliary transformer or emergency diesel Capacity 3500 gpm each - 7000 gpm total *1ee note below Bead (approximately) 435 ft Beat £xchon1era Number Heat load each (See Section 6.2.4.5)

Primary aide flow (containment water)

• 10,700 gpm *aee note below Secondary aide flow (river water) 7,000 gpm *aee note below dP - river water to containment water 20 pd Design temperaturea River water 95*r

• Containment water 165*F Primary '(shell) design presaure 375 pai Secondary (tube) deaign preaaure 375 *pai Beat !xchon1er Code

'nle shell side of the LPCI heat exchanger is conatructed of carbon steel A212, Grade B. 'nle heat exchangers (2 per unit)*were built to ASM!Section III (1965), Class C requirement& as shown on the manufacturer'* apecification sheet. Signed Certificate of Shop Inapection Report* indicate that the heat exchangers were constructed in accordance with the applicable code.

Radiography Requirement* (aee Reference)

GE Specification No. 21A5451 (Rev. 1), Section 4.0 atatea that the exchanger shall be teated in accordance with ASME III, Claes c. The Berlin Chapman

,Specification Sheet atatea that the heat exchanger waa built to Section III.

'Also, the manufacturer'* Data :Sheet givea the ahell joint efficiency of 1001 an.d radiography aa "Complete".

Containmept Spray Syetem Containment Spray Header*

Number 2 Size 8 in. sch. 160

• No. nozzles (each) 160 Type nozzle Fog jet Suppression chamber spray header Number 1 Size 4 in: acb. 40 No. nozzles 12 Type Fog jet

• * 'nle 10,700 and 7000 gpm are design parameters uaed for specification of the LPCI/CCSW heat exchanger. Other flow ratea may be utilized for design basis evaluation* (ref er NFS letter and calculation RSA-D-42-01")<.

ZDFSAR/34 .

ZFSAR92/34/48

. O.l1d ~1~ ~o~* +he. u'f\a..l~?I?

de"-'C'<I W I Y1 Sec-hon 5. ~. ~. 3. I

.* v < -

June 1992 6.2.4-17 CECo'* Nuclear Fuel Services (NFS) analyzed the effect of the lower heat rejection rate on the de*ign basis LCCA. Their report documented tbat tbe mo*t l:lmitina c..e*i* for a situation where a *mall line break on tbe

• Isolation Conden*er render* it inoperable coincident with one LPCI beat exchanger to be out of *ervice. The final analy*i* *how* the increa*e in maximum bulk torwf water temperature to be lea* than 2*r. The re*ultant local temperature (19S*F) is still well below the maximum permi*sible value of 2os*r.

As determined by Perfex, the affect of this modification on flow induced vibration and seismic response will result in a design equal to or *lightly more conservative than the original design. Additionally, AL-6XN'* thennal expansion is clo*e enough to that of the CuHi material *o a* to not cause a warpage problem during the combination of both the AL-6XN and CuNi tube material installed in the affected heat exchanger. The Station Technical Staff will be responsible for creating a new Eddy Current Te*t Standard for the future inspection of the new material.

For schedule -and economic reasons, the tubes will be replaced a1 the old material fails. This will be ongoing taak for many outages until all four (4) heat exchangers are completely retubed with the new material. To avoid

• • holding the.modification package open that long, the modification will be considered "complete" after the first outage that replace* any of the tubes with the new material.

To ensure that other design basis evaluations are not affected by the rt*

replacement of these tubes, the total number of plugged tubes plus tubes().. ~~~

replaced with tbe new material will be limited to 61 ef ~he total heat '<~'((1 0 11a...! \*~

eueheft1e* tttbes. The 61 limit is based on the number of excess tubes 1

Ct%-~'o :\l.°'

provided in the LPCI heat exchanger design. This limit will ensure that the y(c:\\l~ V\~

design basis of heat exchanger capability will not be reduced. u 1vc?<- <

aP~

Based on the above information, the BWRED concludes that AL-6XN can be used ' ,~

to replace the existing CuNi heat exchanger tubing as required on an  ?\#fj J.

"as-needed" basis) in a.ccoY-cla.ric.e. w'H-J.i ~e.. gu1del1nes d..~vc.. ~...a 111 J- (p ~o )(ti-\)~

R~~'(erce. 7,

*6

• 2. 4

• 5

• 5

• 0 . R.EW!ffCES ~p

1. "LPCI Beat Exchanger Ml2-2-86-32 ' 33 Dresden Station", SNED memo M. T. Fredrick to B. E. Bliss, 7/28/87.

2. "Suppression Pool Temperature Limits for BWR Containment",. USNRC NUP.EG-0783, Rev. 1.

3. "P.ETRAN02 Analysis of Suppression Pool Temperature Response at Quad Cities 1/2 and Dresden 2/3", NFSR-0019.

4. "Dresden 2/3 Nuclear Gen<?rating Plant Suppression Pool Temperature Response", NEDC-22170 ~ 7/82..

5. "Suppression Pool Temperature Monitoring System Bulk Temperature Accuracy Assessment for the Dresden 2 ' 3 and Quad Cities l ' 2 Station*"* Nutech report COH-27-210, Rev. O.

6. "RETRAN Computer Code Certification", NFSll-0026, 9/84.

ZDFSAll/34 ZFSAR92/34/68 7 11 ~~om~a:i>C(l !> .+'oY Tc..ibe.. 'Repla.al'Y'lerrt- \JeY-su.s f>f u~~1Ylj

. on I-Per. H-M'.t E;<di.a n~.ers I f ) f\J FS Tr-an.nl'1; tfa.R.. da..~d J..Jo H'VYl her 2 'f 1 ('=Jt:;'Z. I. F?cei::~ -fr, B. V1d /,,

Bxhibit C BNC-QE-06.l Revi*ion 5 Page l of l 10CP'R50~59 Safety Evaluation Cover Sheet Station cD~r~e~s~d~e~n.__~~~~~~~~~~~~~~~~~~~

Hodif icat ion/Minor Plant Change # UFSAR UPQATE Design Issues Worksheets have been completed prior to Safety Evaluation. The following design issues could impact the Safety Evaluation and should be considered during performance of the Safety Evaluation, particularly during Steps 5 (normal operation) and 6 (failure modes):

Ml3, Ml5, Ml6, Ml9, OPl, OPS, R7, 57, S'!'l

( J This evaluation identified an Unreviewed Safety Question. See Item 14 on

- -- *the 10CFR50. 59 Safety Evaluation form.

( J A Technical Specification change is required and a Technical Specification Revision Request has been prepared. See Item 14 on the 10CFR50.59 Safety Evaluation form *

• • (X] This evaluation did not identify an Unreviewed Safety Question and no Technical Specification change is required. The modification or minor plant change may be installed without prior NRC approval.

cz?L~

tognizant Engineer Date !z.//rz_

/ }

Date Design Superintendent or Supervisor

• QE-06.l PECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.1 Revision 5 Page 1 of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUA~IOH

1. List the documents implementing the proposed change.

NA

2. Describe the proposed change and the reason for the change. ___ .

The changes are being incorporated to correct inconsistencies between the UFSAR and the actual equipment/components of the LPCI/CCSW system. The changes are as follows:

1) Provide the Design Basis parameters and results of analysis for Containment Long Term heat up post LOCA.

These results include the resultant peak pool temperature post accident.

2) Provide a revised acceptability for replacement of LPCI heat exchanger tubes with AL-6XN tube material.

3) Provide a table with required NPSH and actual NPSH for the LPCI pumps under the analyzed conditions and parameters.

3. Is the change:

(X) Permanent

( ) Temporary -

Expected duration --------------------------------~

AND Plant Mode(s) restrictions while installed ----------------------------~

(NONE if no plant mode restrictions apply)

4. List the SAR sections which describe the affected systems, structures, or components (SSCs) or activities. Also list the SAR accident analysis sect*ions which discuss the affected sscs or their operation. List any other controlling documents such as SERs, previous modifications or Safety Evaluations, etc.

UFSAR Sections 5.2, 6.2 and 14.2 SER 104301 50.59 Safety Evaluation for previous UFSAR change on LPCI Heat Exchanger Tube Replacement dated April 7, 1992

• QE-06.1 CECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.l Revision 5 Page 2 of 9 station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION

s. Describe how the change will affect plant operation when the changed SSCs function as intended (i!e., focus on system operation/interactions in the absence of equipment failures). Consider all applicable operating modes. Include a discussion of any changed interactions with other sscs.

The changes being made to the UFSAR will not affect plant op~ration. The Tech Spec surveillance limits for the LPCI and ccsw pumps are unchanged by these changes to the UFSAR. The changes consist of the following:

1) Updates to Section 6.2 which provide clarifications on the

• . LPCI/CCSW Pump flows and the heat exchanger duty.

2) Updates to Section 5.2.3.3 which provide the Bases and results of the long term containment heat up analysis post LOCA. *

3) Updates to Section 6.2.4.5.1.0 to provide conditions under which tube repla9ements with AL-6XN tube material may be performed.

6. Describe how the change will affect equipment failures. In particular, describe any new failure modes and their impact during all applicable operating modes.

The descriptive changes will not affect any equipment failures.

The analysis was performed to verify that the existing equipment will satisfy the requirements of the Design Basis Accident (OBA).

7. Identify each accident or anticipated transient (i.e., large/small break LOCA, loss of load, turbine missiles, fire, .flooding) described in the SAR where any of the following is true:

The change alters the initial conditions used in the SAR analysis

~The changed SSC is explicitly or implicitly assumed to function during or after the accident operation or failure of the changed SSC could lead to the accident ACCIDENT SAR SECTION LOCA 14.2

• . QE-06.l CECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.1 Station/Unit *=D=r=e=s=d~e~n.__

Revision S Page 3 of 9

_______________________________________ / ______

Exhibit B lOC~S0.59 SAFETY EVALUATION

8. List each Technical Specification (Safety Limit, Limiting Safety System Setting or Limiting Condition for Operation) where the requirement, associated action items, associated surveillances, or bases may be affected. To determine the factors affecting the ~pacification, it is necessary to review the FSAR and SER where the bases section of the Technical Specifications does not explicitely state the basis.

SECTIONS -3. 5 / 4. 5, 3. 7 / 4. 7

9. Will the change involve a Technical Specification revision?

• [

• J Yes * [ X J No If a Technical Specification revision is involved, the change cannot be implemented until the NRC issues a license amendment. When coapleting Step 14, indicate that a Technical Specification revision is required *

• QE-06.1 DECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.1 Revision 5 Page 4 of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION

10. To determine if the probability or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR may be increased, use one copy of this page to answer the following questions for each a~cident listed in Step 7. Provide the rationale for all NO answers.

Affected accident =L~O~C~A.:....~~~~~~~~-

SAR Section: 14.2 May the probability of the accident be increased? [ ) Yes (X] No The updates to the UFSAR have no affect of the porbability of the accident because no physical changes are being made to any equipment or systems.

May the consequences of the accident (off-site dose) ( ) Yes (X] No be increased?

The analysis has verified that the revised para~eters provide the same level of accident mitigation as originally designed.

May the probability of a malfunction of equipment ( ) Yes (X] No important to safety increase?

The changes are being made to the Design Basis and no equipment changes are being made, therfore the probability of equipment failure remains unchanged.

May the consequences of a malfunction of equipment ( ) Yes [X) No important to safety increase?

The accident mitigation capability fo the Containment System is unchanged from the original design analysis. The analysis validates the capability of the exisitng equipment to perform its original design function.

If any answer to Question 10 is YES, then an Unreviewed Safety Question exists *

• QE-06.1 DECA version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.l Revision S Page S of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION

11. Based on your answers to Qu~stions 5 and 6, does the change adversely impact systems or functicns so as to create the possibility of an accident or malfunction of- a type different from those evaluated in the SAR?

[ ] Yes [X) No Describe the rationale for your answer.

The analys.is validates the ability of existing LPCI/CCSW system components to perform their original design functions. No

• physical*equipment changes have been made, therefore is no possibility of an unanlyzed accident occurring.

If the answer to *ouestion 11 is Yes, then an Unreviewed Safety Question exists *

• QE-06.l CECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.l Revision 5 Page 6 of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION

12. Determine if parameters used to establish the Technical Specification limits are changed. Use one copy of this page to answer the following questions for each Technical Specification listed in Step 8. List the Technical Specification, Technical Specification Bases, SAR and SER Sections reviewed for this evaluation. __________________________________________

TECH SPEC 3.5. 3.7 4.5, 4.7 UFSAR SECTIONS 5.2 AND 6~2 SER 104301 Evaluation of Technical Specification (Enter N/A' if none are affected and check last option.)

N A (Check appropriate condition):

[ J All changes to the parameters or conditions used to establish the Technical Specification requirements are in a conservative

. direction. Therefore, the actual acceptance limit need not be identified to determine that no reduction in margin of safety exists - proceed to Question 13.

[ ] The Technical Specification or SAR provides a margin of safety or acceptance limit for the applicable parameter or condition. List the limit.(s)/margin(s) and applicable reference for the margin of safety below - proceed to question 13.

[ ] The applicable parameter or condition change is in a potentially non-conservative direction and neither the Technical Specification, the SAR, or the SER provides a margin of safety or an acceptance

• • limit. Request Nuclear Licensing assistance to identify the acceptance limit/margin for the Margin of Safety determination by consulting the NRC, SAR, SER's or other appropriate references.

List the agreed limit(s)/margin(s) below.

[XJ The change does not affect any parameters upon which Technical Specifications are based; therefore, there is no reduction in the margin of safety. Proceed to question 14.

List Acceptance Limit(s)/Margin(s) of Safety QE-06.1 DBCA Version 2.0A

Exhibit E Mod # UFSAB UPPATE ENC-QE-06.l Revision 5 Page 7 of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION

13. Use the above limits to determine if the margin of safety is reduced (i.e., the new values exceed the acceptance limits). Describe the rationale for your determination.. Include a description of compensating factors used to reach that conclusion.

If a Margin of Safety is reduced an Unreviewed Safety Question exists.

QE-06.1 CECA Version 2.0A

Exhibit E Mod # UFSAR UPDATE ENC-QE-06.l Revision 5 Station/Unit ~D~r~e~s~d~e~n.__ _______________________________________

Page 8 of 9

/~~---

Exhibit E 10CFRS0.59 SAFETY EVALUATION

14. Check one of the following:

( ) An Unreviewed Safety Question was identified in Step 10, Step 11, or Step 13. The proposed change MUST NOT be implemented without NRC approval.

(XJ No Unreviewed Safety Question will result ( Steps 10, 11, and 13)

AND no Technical Specification revision will be involved. The change may be implemented in accordance with applicable procedures.

[ A Technical Specification revision is involved; but no Unreviewed Safety* Question will result. The proposed change requires a License Amendment. Notify Station Regulatory Assurance and Nuclear Licensing that a Technical Specification revision is required.

Mark below as applicable.

[ ) The change is not a plant modification or minor plant change and will not be implemented under. 10CFR50.59. Upon receipt of

.. the approved Technical Specification change from the NRC, the change may be implemented.

[ ] The change is a plant modification or minor plant change.

Mark below as applicable.

A revision to an existing Technical Specification is required. The change MUST NOT be installed until receipt

• of an approved Technical Specification revision.

The change will not conflict with any existing Technical Specifications and only new Technical Specifications are required. In these cases, Nuclear Licensing may authorize installation, but not operation, prior to receipt of NRC approval of the License Amendment. If such authorization is granted, the block below should be checked.

[ ) Nuclear Licensing has authorized installation, but not operation, prior to receipt of NRC approval of the License Amendment. The 10CFR50.59 Safety Evaluation indicates that no Unreviewed Safety Question will result and provides authority for installation only~

QE-06.l DECA Version 2.0A

Exhibit E Mod # UFSAB UPDATE ENC-QE-06.l Revision 5 Page 9 of 9 Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION Note: Partial Modifications and/or separate 10CFRS0.59 reviews for rtions of the work may be used to facilitate installation.

J Date

15. The reviewer has determined that the documentation is adequate to

. sup~~ove con~u~ion(\and agrees with the conclusion.

RevJ.ewer _ij_~- \ ~

(Design Superintendent/Supervisor) 1 Y.AZ

---

7..,..... . . D_a_t_e QE-06.l DECA Version 2.0A

• • Ex.

ENC*QE-06.1 Revision 5 Page 1 of 17

• DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE ELECTRICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION safety related electrical or NO THE CHANGE IS TO THE UFSAR ONLY, NO EQUIPMENT CHANGES ARE BEING E1 Is Class 1E equipment involved? l&C system, basis PERFORMED described in design input doc1.111ent E2 Is there any potential for *separation of voltage NO THE CHANGE IS TO THE UFSAR ONLY, NO EQUIPMENT CHANGES ARE BEING control and power circuit classes, induction effects PERFORMED interaction? on control signals E 3 Has a sneak circuit analysis potential shorts, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE been c~ieted7 inadvertent connections, BEING MADE i.nintended operating mode E 4 Is redundancy of existing backup of protection NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE

_systems reduced or system, fire zone BEING MADE c~romised? consideration, independent control station, interconnection of r~t system, power supply crossties E5 Are safety related circuits buffer ~lifiers, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE Isolated and separated from automatic switchgear, BEING MADE non-safety related circuits? separate cable rl61s, electrical and physical separation E 6 Is safety related (Class 1E) bus capacity, automatic NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE bus Integrity aiaintained7 Isolation, load shedding BEING MADE E7 Has diesel generator or overload potential, load NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE battery loading been sequencing and shedding, BEING.MADE checked? i.nlnterruptible power E 8 Are there adequate fail safe automatic transfer, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES AllE protection features for both r~t systems, failure BEING MADE coqxinents and systems? lllOde status

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue changes the nonaal operetion or the failure lllOdea/effects resulting from the modification.

QE-06.1

• DECA Version 2.0A

• DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE Ex ENC*QE *06. 1 Revision 5 Page 2 of 17 ELECTRICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION '

E 9 Does the design provide minimize extent of outage, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CllAMGES ARE fault trip coordination on interaction_with load BEING MADE the system and interfacing shedding, operations systems? sequencing, timing E 10 Is actuation time of response time, reactor trip NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE protection devices and time, containment BEING MADE circuitry cocrpatible with all isolation, interaction with requirements? other systems E 11 Are in-service periodic availability for testing, NO testing and inspection of frequency of testing, CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CllAMGES ARE system performance potential for undesirable BEING MADE addressed? side effects E 12 Does the modification of control panel layout, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE control panels incorporate control f~tion, separate BEING MADE h~n factors objectives? evaluation, control room (h~n factors requires a panels and remote panels separate evaluation)

E 13 Has bypass and inoperable verification of status, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE status indication of Class 1E technical specification BEING MADE protection equipment been c~llance, operational included in the design? requirement E 14 Does the design adequately new off-site sources, new NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE address Radio Frequency electrical or electronic BEING MADE Interference (Rfl) and equipnent, new on-site Electromagnetic coanaJnication devices, Interference CEMI)? hand-held radio signals E 15 Do system logic logic diagram, lnstrunent NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE configuration changes alter loop dlagr* BEING MADE system design?

• list this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue chengea the normel operation or the failure lllOdes/effects resulting from the lllodiflcatlon.

QE*06. 1

• DECA Version Z.OA

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ENC-Qf-06.1 Revision 5 Page 3 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE ELECTRICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION E 16 Are there any grounding equipment ground, ground NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE changes or requirements? grid, disconriecting a BEING MADE ground E 17 Have Control Room Panel equipment cha"9es, i~ct NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE additions and deletions been on seismic qualification of BEING MADE revised for seismic panel, panel requal ification qualification iq>act?

E 18 Are there any other NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE Electrical or l&C Issues BEING MADE that should be addressed?

If so, list and discuss them here.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue chenges the nonaal operation or the failure modes/effects resulting from the modification.

QE-06.1

• DECA Version 2.0A

DESIGN ISSUES WORKSHEETS Hod #UFSAR UPDATE

..9 ENC*QE-06. 1 Revision 5 Page 4 of 17 FIRE PROTECTION ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION F1 Have all ignition sources hydrogen in contairment NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE

.been adequately controlled? arcing contacts, static BEING MADE electric charges, open flames, off-gas control F 2 Do any additional sources of contiustibles; materials NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL.CHANGES ARE energy cause the capacity to that could react to BEING MADE a fire zone to be exceeded? produce contiustible gas, Zn or Al in contairment F3 Are all materials of excessive propagation rate, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE construction appropriate for controlled materials, BEING MADE fire protection purposes? radiation effects, potential for failure in a fire F4 Is there additional storage electrical insulation NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE of contiustible material or coatings, gas supplies, BEING MADE have contiustible materials additional cable trays been added as part of constitute added fire lllOdi f !cation? loading F5 Are there any new potential *holes through fire walls or NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE paths for fire propagation or stops, ducts, daq>er BEING MADE crossing of fire zone failure mode bowldaries7 F6 Have changes coq>romlsed thermal insulation or NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CIWIGES ARE testing or inspection of the shielding which could BEING MADE fire protection system? block access F7 Have any changes been new failure lllOdes, move NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CIWIGES AIE lll8de that degrade required or penetrate fire walls, BEING MADE fire detection, control or reduce capacity of water protection? supply system, tie-In to fire detection system

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if.the Issue changes the noMaal operation or the failure lllOdes/effects resulting from the lllOdiflcatlon.

QE-06.1 . OECA Veralon 2.0A

• DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE Ex B ENC*OE-06.1 Revision 5 Page 5 of 17 FIRE PROTECTION ISSUES I

No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION F8 Are there any other Fire NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE Protection Issues that BEING ~E should be addressed? If so, list and discuss here.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes *the nonnal operation or the failure llOdes/effects resulting frora the modification.

QE-06.1

• OECA Version 2.0A

• 1

hhiblt B ENC-QE-06.1 Revision 5 Page 6 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE FLOODING ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION fL 1 Is there any increase in the circulating water, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE potential for internal condenser, D~6 pipe lines, BEING MADE flooding? Suppression J>ool, Fan coolers, Se~vice ~ater heat exchangers, Drywell chillers*, Sprinklers, failed check valves, augunented fire protection systems Fl 2 Are any areas or equipment Lower levels, ~atertight NO *CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE susceptible to flood roocns, Electrical BEING MADE damage? equipment close to floor, P~s, Motors, Air C~ressors, Electrical Buses, Breakers, direct or Indirect failure Fl 3 Are any potential paths for Holes through walls, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE flood propagation created? floors, & doors designed BEING MADE to be watertight, Floor Drains, Ventilation Ducts, backflow, siphoning, site~

topography Fl 4 Is the capability to isolate extended removal or NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE

.or cope with flooding disengagement of valves, BEING MADE reduced? ~ elanns, indicators, s~llng systems, opening or isolating pipeline, blocking or closing drains, s~.

FL 5 Are there adequate design leak protection or isolation NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE considerations to mitigate devices drainage systems, BEING MADE flooding? barriers, separation of equipment

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes the nonaal operation or the feilure lllOdes/effects resulting from the modification.

QE-06.1 DECA Version 2.0A

E B ENC*QE*06.1 Revision 5 Page 7 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE FLOODING ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION '

FL 6 Are there any other Flood NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE Protection Issues that BEING MADE should be addressed? If so, list and discuss here.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes the non111l operation or the failure lllOdes/effecta reaultl~ from the modification.

QE*06.1 . DlCA veralon Z.QA

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ENC-QE-06. 1 Revision 5 Page 8 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE MECHANICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION M1 Are any high energy lines jet iQ'1ingement, pipe NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE added or affected? whip, special supports BEiNG MADE M2 Is the vulnerability to new missile source(s), NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE internally generated missiles ~ rotor breakup, valve . BEING MADE increased? stem ejection, pressure vessel appendages, change In missile protection requirement M3 Is the vulnerability to tornado driven object, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE externally generated missiles airplane, protection for BEING MADE increased? . new facilities, change in missile protection requirement M4 Is there a potential for loose cleanliness requirements, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE particles within piping heat exchanger plugging, BEING MADE systems or c~nts? If effect on in-line devices so, how is it addressed?

M S Could deformation or equipment Sl4lP0rt failure NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE catastrophic failure ill'18ir results in de&radation of BEING MADE the safety fl.Sletlon of the safety system directly or system, c~ts or indirectly, over structures being lllOdified, pressurization failure, or other surrO\a'lding safety excessive flow forces on related systems? valve stem causing 111i soperat I on M6 Is the safety classification lllOdif I cat I on cif NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE of modified systems Interconnecting systems, BEING MADE consistent with and change from non-safety appropriate for the safety related to safety related at classification of existing contalnnent penetration, systems? support attachment point, c~tlblllty of appendages

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the laaua changes the noraal operation or .the failure llOdes/effecta resulting frOlll the lllOdlflcatlon.

OE-06.1

• DECA Version 2.0A

DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE EM-ENC-QE *06. 1 Revision 5 Page 9 of 17 MECHANICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION M7 Is de>\ble valve isolation contairvnent jsolation NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE used if changes frocn class valves, safety classification BEING MADE 1 to any other class or change within a piping non-class portions of a system system, or when a system Is in direct contact with contalrvnent atmosphere?

Is a single valve isolation used in changes from class 2 to class 3, class 2 to non-class, or class three to non-class portions of a system?

M8 Does the system have the fail open, fail close, or fail NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE required fail safe as is at both the BEING MADE protection? Is the safety coq>e>nent s f~tion of the interfacing safety systems preserved 1..f>OO failure?

M9 Is the redundancy of existing backup system for NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAl CHANGES ARE systems reduced by redundancy, adequate BEING MADE Inadequate reliability? reliability designed in for proper redwldancy M 10 11 there an envlronnental certified to operate In a NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE qualification requirement? specified t~rature, BEING MADE (envlronnental qualification hunldlty, and radiation requires a separate envlrOn111ent; by test, by evaluation) verification analysis, or a cOlllbl natl on

" ,, Are there any changes to the environnental profile of high energy l lne routing, changes In process NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES AlE BEING MADE an envlrorcnental per-ters qualification zone?

• Llat this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue changes the no,...l operation or the failure modes/effects resulting from the lllOdlflcatlon.

QE*06. 1 . DECA Venton 2.DA

• DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE

.9.

ENC*Q£*06.1 Revision 5 Page 10 of 17 *

• MECHANICAL ISSUES No.* DESIGN ISSUE KEY W*ORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION M 12 Is seismic qualification maintain structural NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAMGES ARE required? integrity; operate during BEING MADE and after seismic event; category II over category M 13 Have all appropriate design hydrodynamic loads, pipe YES THE REANALYSIS WAS PERFORMED TO VERIFY THE ADEQUACY OF JHE LPCI SYSTEM loads (new and existing) in break loads, thermal loads WITH REVISED CAPABILITIES OF THE HEAT EXCHANGERS. ALL LOADS USED WERE addition to seismic loads VERIFIED AS APPROPRIATE BEFORE COMPLETION Of THE ANALYSIS.

been identified?

M 14 Has the c~tibility of material considerations, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAMGES ARE 111aterials been evaluated? prohibited materials, BEING MADE sealants, coatings, insulation, effect of radiation, erosion/corrosion resistance, containnent restrictions on some materials, stainless/non* stainless interfaces 111isoperation M 15 Have changes been made excessive pressure loss in YES THE ANALYSIS PRCX>UCED POTENTIAL INCREASED TEMPERATURES FOR THE TORUS that could affect the NPSH suction piping, cavitation, WATER WHICH IS THE SUCTION SCJURCE FOR THE LPCI Pt.tlPS. THE REQUIRED AMO for any ~? fluid temperature change. ACTUAL NPSHs FOR THE PUMPS HAS BEEN CALCULATED TO VERIFY ACCEPTABILITY.

M 16 Ara there any changes in balance of flows, YES THE ANALYSIS USED CHANGED PARAMETERS FOR THE HEAT REMOVAL CAPAlllLITIES process parameters? t~rature, pressure Of THE LPCI HEAT EXCHANGERS BASED ON REDUCED FLOWS THROUGH THE IOI FROM limitation of existing BOTH LPCI AND CCSW. THESE PARAMETERS WERE INDEPENDENTLY VERIFIED PRIOR system capability, i~ct TO USE IN THE ANALYSIS.

on design function

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue changes the non11Bl operation or the failure lllOdes/effects resulting frOlll the lllOdlflcatlon.

QE-06.1 - DECA Veralan Z.OA

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ENC*OE *06. 1 Revision 5 Page 11 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE MECHANICAL ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION M 17 Valve Performance as it valve, containnent isolation NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE relates to system function: valves, valve BEING MADE can the valve be placed orientation/configuration, and maintained in the Design Basis Event, valve appropriate position for closure time; isolation logic normal system operation, changes abnormal system operation, and testing mode?

If the valve is a primary containment isolation valve, can it be closed (if necessary) during the long term phase of a Design Basis Event (DBE)?

M 18 Have short*term and long*term containment isolation NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE containment isolation BEING MADE requirements been satisfied?

M 19 Have the rules for single failure criteria YES THE ANALYSIS USES THE LIMITING CASE OF PlMP AVAILABILITY BASED ON A single failure criteria LOCA/LOOP SCENARIO.

been applied correctly?

" 20 Are there any other NO NONE Mechanical Issues that should be addressed? If so, list and discuss here.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes the nol"lll8l operation or the failure lllOdes/effects resulting frOlll the modification.

GE*06.1 OECA Version 2.0A

Exti B ENC*OE-06.1 Revision 5 Page 12 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE OPERATIONAL ISSUES.

No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION OP 1 Will the ope rat .i ng teq>erature, pressure, YES THE OPERATING PARAMETERS AND TECHNICAL SPECIFICATION VALUES FOR THE LPCI conditions of this or any flow, cooling water supply, AND CCSW PUMP SURVEILLANCES HAVE BEEN VALIDATED BY THIS ANALYSIS.

other system be changed? electrical power interr~tions OP 2 Will the operation of any shared source of power NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PH.YSICAL CHANGES ARE other system have any effect system fluid, interlocks, BEING MADE on the system being emergency power modified? priorities OP 3 Will the change have any failure modes, reduction in NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE iq>act on adjacent systems? availability or reliability BEING MADE OP 4 Can the change affect the shared systems, cascading .NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE operation of another system effect, ripple effect BEING MADE indirectly?

OP 5 Has the i~ct on surveillance, operability YES THE TECH SPEC SURVEILLANCE VALUES FOR LPCI AND CCSW Pt.ICP PERFORMANCE operability tests been test, channel check, HAVE BEEN VALIDATED BY THIS ANALYSIS.

considered? *cal lbratlon OP 6 Are there any other NO NONE Operational Interaction Issues that should be addressed? If so, list and discuss them here.

• List this Item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue chengea the no1'1119l operation or the failure IBOdes/effects resulting frOlll the modification.

OE-06.1

• DECA. Version 2.0A

Exli B ENC*QE-06.1 Revision 5 Page 13 of 17 DESIGN ISSUES WORKSHEETS Hod #UFSAR UPDATE RADIOLOGICAL ISSUES_

No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION R1 Are there any changes that wet HEPA filters, cross* NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE affect the engineered safety connection, bypass or BEING MADE feature ventilation system? leakage R2 Are there any changes to high filter pressure drop, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE the controlled leakage backup through air BEING MADE systems CBWR), such as a Intakes, strilctural integrity change in back pressure?

R3 Are stores of persOl'Vlel emergency air supplies for NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE protective equipment control roocn personnel, BEING MADE preserved? emergency breathing air suppt'ies, i~ired access R4 Are there any effects on false readings due to NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE radiation detection and placement, unintended BEING MADE monitoring or alarm shielding, side effects of systems? enclosures R5 Are there any effects on reliability, operability, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE containnent isolation access, containnent spra_y BEING MADE systems, ventilation systems system, iodine removal or contairwient cleanup system?

R6 Has separation or secondary side detection NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE primary/secondary coolant system, equlpnent BEING MADE systems (PWR> or leakage, bcK.rdary changes contalnnent drywell (BWR) been maintained?

R7 Are there any effects on contalnnent spray

• NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE fission product control for cleanup system BEING MADE Incidents/accident or post accident cleanup and monitor points?

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes the normal operation or the failure lllOdes/effects resulting from the lllOdification.

QE-06.1

• DECA Ver1lon 2.0A

• DESIGN ISSUES WORKSHEETS Hod #UFSAR UPDATE hhl Revision 5 8

ENC*QE-06. 1 Page 14 of 17 RADIOLOGICAL ISSUES*

No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION R8 Have *adequate provisions monitoring required, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE been made to ~ontrol human error.protection, BEING MADE effluent containnent levels? potential releases, s~

cont Biii nation R9 Is there any potential for decontamination, ALARA, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL *CHAllGES ARE additional radiation recb:tion in shielding BEING MADE exposure?

R 10 Are there any other NO NONE Radiological Issues that should be addressed? If so, list and discuss here.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the issue changes the no1'9111l operation or the failure lllOdes/effects resulting frOll the lllOdiflcatlon.

QE-06.1 DECA Version 2.0A

E;x ENC-QE-06. 1 Revision 5 Page 15 of 17 DESIGN ISSUES WORKSHEETS Mod #UFSAR UPDATE SITE RELATED ISSUES I

Ho.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION s 1 Is there any change in the Change the fence line, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE exclusion area or site construct a new building BEING MADE boundary conditions which containing ~adioactive would increase the on-site materials, relocate or off-site dose rates? activated materials.

• ~

s 2 Is the site radioactive quantity or c~sition of NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE material inventory control radioactive materials on BEING MADE affected? site - increased or changed s 3 Are release and dispersion stack height change, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE of effluents affected? concentration of radwaste, BEING MADE or other factors affecting effluent pathways, contairment isolation valve leak rates or closure times s4 Are there any changes failure effects of non- NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE affecting protection of safety safety related structure or BEING MADE class structures from natural syste11, change to surface phenomena and water control structures, meteorological conditions secondary effects (tornados, rain loads, snow loads)?

s 5 Are there any potential placing equipment In close NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE effects on security barriers proximity to guardhouse BEING MADE or controlled access? or security equipment s6 Are any potential hazards fire source, explosive NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE added to the site or material, toxic material, BEING MADE exclusion area? radwaste material, on-site or off-site, permanent or teq>e>rary.

s7 Are there any changes to quantity, t~rature, YES THE REVISED PARAMETERS USED IN THE ANALYSIS ACaJMMll>ATE THE REDUCED cool! ng water S'-'3Pl y aedi..nt content, aquatic CAPABILITY OF THE LPCI HEAT EXCHANGERS FOR HEAT IEJIJVAL AT REDUCED CCSW capacity or characteristics? growth potential, flowrates, Flo.IS.

~ curve changes, etc.

• List this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue changes the normal operation or the failure modes/effects resulting from the lllOdlflcatlon.

QE-06.1

• DECA Version 2.0A

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Exli B ENC*Q£*06.1 Revision 5 Page 16 of 17 DESIGN ISSUES WORKSHEET~ Mod #UPSAR UPDATE SITE RELATED ISSUES No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS POR CONCLUSION s 8 ls the stability of subsurfac~ ground water level, soil NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE materials or foundations for ph, soil response to BEING MADE Class 1 structures affec~ed excitation, excavating near directly or indirectly? existing structures, subsidence s 9 Is plant access altered or roadway or railroad NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE affected? changes, GSEP, access BEING MADE gate change, underground tl.l'INll s 10 Yill site topography changes excavation, topography NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE increase the potential for BEING MADE external flooding?

.s 11 Are there any other Site NO NONE Related Issues that should be addressed? If so, list and discuss here.

• List this lte11 on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue changes the non111l operation or the failure modes/effects resulting frOll the llOdlflcatlon.

QE*06.1 . DECA Version Z.OA

Revision 5 Page 17 of 17 DESIGN ISSUES WORKSHEETS Hod #UFSAR UPDATE STRUCTURAL ISSUES -

No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT? PROVIDE BASIS FOR CONCLUSION ST 1 What is the seismic Category I or non-seismic YES THE CONTAINMENT STRUCTURE IS SEISMIC CATEGORY I. THE LOCA ANALYSIS Oii classification of the LONG TERM SUPPRESSION POOL HEAT UP IS NOT DEPENDENT Oii THE SEISMIC structure? QUALIFICATION OF THE STRUCTURE.

ST 2 Is the response subsystem analysis, NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE characteristic of the fln:tamental frequenc.,*, BEING MADE existing structure changed stiffness, coupling, adding by the modification? or redistributing mass ST 3 Does the modification enlarge openings, create NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE degrade the structure nunerous discontinuities, BEING MADE integrity of the existing additional loads, structure? penetrations, cuaulative effects ST 4 Does the modification create Seismic II over I, non- NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE the possibility of failure seiSllic/non*safety BEING MADE

• due to failure of non-seismic structures or equipment equipment affecting nearby seismic category I equipment?

ST 5 Are there any changes that obstruct surface, reduce NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMEllT/PLANT PHYSICAL CHANGES ARE would affect testing ard/or availability for testing, BEING MADE in-service inspection of the restrict access structure?

ST 6 Has qualification by testing, purchase of seismically NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE as opposed to analysis, been qualified structures or BEING MADE considered for seismic c~ts, size limit, structures or COlllpOnef'lts? weight ll*lt ST 7 Are there any other 110 NONE structural issues that should be eddressed? If so, list and discuss here.

• list this item on the 10CFR50.59 Safety Evaluation Cover Sheet if the Issue chenges the nol'll8l operation or the failure lllOdes/effects resulting from the llOdlflcatlon.

QE-06.1 - DECA Veralon 2.0A

• '°"1

November 30, 1992 Cl1ROIJ # l'i'f no To: C. Schroeder

Subject:

Post DBA-LOCA LPCI NPSHA Evaluation

References:

1. Nuclear Engineering Department calculation NED-M-MSD-43, "Dresden LPCI Pumps NPSHA Evaluation Post DBA-LOCA".

dated November 30, 1992.

2. General Electric Report No. GENE-770-26-1092 "Dresden Nuclear Power Station Units 2 & 3 LPCl/Containment Cooling System Evaluation", November, 1992.

Post DBA-LOCA torus conditions were determined by GE in Reference 2 and were used to calculate the available NPSH for the LPCI pumps at Dresden Station (Reference 1). The results (Table 1) indicate that the available NPSH is greater than the NPSH required (with

• margin) for all four cases analyzed in Reference 2, and therefore adequate to protect the pump under these conditions.

If there are any questions or comment, please*contact Harry Palas at x7494.

prepared by:_* ~-i---.

....._P_al_as approved cc: S. Eldridge R. Kolflat

Calculation No. NED-M-MSD-43 Dresden LPCI Pumps NPSHA Evaluation- Post DBA-LOCA lotal Single Torus Specific Vapor Suction Flow Pump Torus Pressure Static Volume Pressure Piping NPSHA NPSHR Margin Case (gpm) Flow (gpm) Temp (F) (psia) Head (ft) (ft3/lb) (psia) Losses (ft) (ft) (ft) (ft) 3 10000 5000 168 18.7 13.32 0.01644 5.7223 4.72 39.32 30.00 9.32 3A 8916 4458 171 19.1 ' 13.32 0.016457 6.1318 3.75 40.30 26.90 13.40 4 5000 5000 180 19.9 13.32 0.01651 7.511 3.77 39.00 30.00 9.00 4A 3881 3881 186 20.6 13.32 0.016547 8.568 2.27 39.72 25.70 14.02 TABLE I

(

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COMMONWEALTH EDISON COMPANY TITLE PAGE.

~ED- M - MSC - 'f'3 . PAGE CALCULATION NO. l OF. 7

_tsa SAFETY RELATED 0 NON-SAFETY RELATED CALCYLATION TITLE D~s~ LPc1 Pu""'ps NPSHA .vo.luo:f,'"n.

Pc~ DBA 0GA EQUIP NUMBER(S) STATION/UNIT SYSTEM

~(:,) - t5o~ A/e/tjD Df!rJ.en ~ g.3 l-/'CJ REV. CHRON # PREPARER

• DATE REVIEWER DATE APPROVER DATE Ql41.0 UHlllTI lllV. l "-~------------........- - - - - - - - - - - - - - - -

Calculation Ho. NED-M-KSD-43 Dresden LPCI Pumps NPSHA EValuation - Post DBA-LOCA Purpose/Obiective; Calculate the Net Positive suction Head Available (NPSHA) for the LPCI pumps at Dresden Station under post-accident conditions as outlined in Reference 2, and compare with NPSH required (NPSHR) to ensure pump protection.

Assumptions/Inputs; The NPSHA is calculated for each of the four cases anal¥zed by General Electric in Reference 2. Inputs to this calculation were taken from Tables 3, 4 and B.2 of Reference 2 and are summarized in Table l below:

Reduced LPCI Total Maximum Suppression Pumps Flow Suppression Chamber Case /Loop (gpm) Pool Temp(F) Pressure(psia) 3 2 10000 168 18.7 3A 2 8916 171 19.l 4 l 5000 180 19.9 4A l 3881 186 20.6 Table 1 These calculations include the followinq assumptions:

l) An even split of flow is assumed between two pumps operatinq in parallel.

2) Suction pipinq losses based on calculations in References l and 5.

J) NPSHR values taken from Reference l (Table 2 - no temperature correction). For cases 3A and 4A, NPSHR values were obtained throuqh linear interpolation~

References; l) R. Kolflat letter report titled "Alternate Shutdown Coolinq Core Spray and LPCI pumps", Chron #841425 dated April 23, 1984

2) General Electric Report No. GENE-770-26-1092 "Dresden Nuclear Power Station Units*2 & 3 LPCI/Containment Coolinq System Evaluation," November, 1992

3) S. Eldridqe letter to c. Schroeder titled "Submerqence of LPCI Discharge Line Post LOCA Dresden Units 2 and 3" dated September 29, 1992, chron# 0115532

4) ASME Steam Tables, 1967

5) Alternate Shutdown coolinq Core Spray and LPCI pump notes and back-up calculations for Reference 1, R. Kolflat, circa 4/89

Calculation No. HED-K-KSD-43 Dresden LPCI PUmps HPSllA Evaluation - Post DBA-LOCA Eauations:

• Net Positiv~ SUctioQ.Head Available (NPSHA) is determined using the following equation (Reference 1):

NPSHA (ft)

= Torus Pressure +

Static Head Vapor Pressure Suction Losses

( 1) where: Torus Pressure= given in Table 1 (psia); converted to feet using specific volume Static Bead * = the minimum water elevation expected above the LPCI pump suction as calculated below:

Minimum Torus water level elevation 491. 5' (includin9 maximum post-LOCA draw down as discussed in Reference 3)

LPCI pump suction elevation 478.13' Static Head 13.32' Vapor Pressure = from Reference 4, in psi a;. converted to feet using specific volume Suction Losses = pipinj losses in feet *

• K

• Q ' K calculated at Q = sooo gpm

• LPCJ N'PSHA calculations; using suction losses from References 1 and 5. (Tables 2 and 3)

Using Equation 1 and the inputs provided above, the NPSHA is calculated for each of the four cases (Table 4). The required NPSH is also provided and the difference between the two is calculated.

• summary/Conclusions:

Post DBA-LOCA torus conditions were determined in Reference 2 and were used to calculate the available NPSH for the LPCI*

pumps -~t Dresden Station. The results in Table 4 indicate that the available NPSH is greater than the NPSH required (with margin) for all four cases, and therefore adequate to protect the pump under these conditions *

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CaJculatlon No. NED-M-MSD-43 Dresden LPCI Pumps NPSHA Evaluation- Post CBA-LOCA

-Total SingJe Torus Specific Vapor Suction Flow Pump

• Torus Pressure Static Volume Pressure Piping NP SHA NPSHR Margin Case (gpm) Flow (gpm) Temp(F) (psia) Head (ft) (ft3/lb) (psla) Losses (ft) (ft) (ft) (ft) 3 10000 5000 168 18.7 13.32 0.01644 5.7223 4.72 39.32 30.00 9.32 3A 8916 4458 171 19.1 13.32 0.016457 6.1318 3.75 40.30 26.90 13.40 4 5000 5000 180 19.9 13.32 0.01651 7.511 3.n 39.00 30.00 9.00 4A 3881 3881 186 20.6 13.32 0.016547 8.568 2.27 39.72 25.70 14.02

COMMONWEALTH EDISON COMPANY REVIEW CHECKLIST CALCULATION NO: l\/CD -M _ M~D-'f3 REV. 0 PAGE 7 OF 7 DATE:

m tQ

~ 0 1. IS ntE oeJECTM OF TME ANALYSIS a.EARLY STA1ED7

' 0 2. ARE ASSUIWT10NS AlllJ ENOMEUNG JUDGEMENTS VAUD AND DOCUMENTED?

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c

3. ARE ntDll AUUWT10Na 'IMAT NEED ViRFICATION7

4. ARE ntE REFEAENC:O (LL DRAWINGS. CODES. STANDAADll US'1'U> IV REVISION EDITION. DAlE. ET'C.1

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C:OOU. STANDAADI. AND RICI. CIUIJll7

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~~.) '----------------------------------------------------------------------------------

COMMONWEALTH EDISON COMPANY TITLE PAGE CALCULATION NO.

• NED-M-MSD-43 I PAGE 1 OF 13

~ SAFETY RELATED NON-SAFETY RELATED CALCULATION TITLE Dresden LPCI/Core Spray Pumps NP SHA Evaluation Post DBA-LOCA L,

EQUIP NUMBER{S) STATION/UNIT SYSTEM 2 {3) - . 1502A'/B/C/D Dresden 2 & 3 LPCI/Core Spray 2(3) - 1401A/B REV. CHRON # PREPARER DATE REVIEWER DATE APPROVER DATE 0 194745 H. Palas iY'o/4f:2..

• R. Kolflat 1y'.10/<t~ P. Dietz 1Y.J~.z.,_

1 198391 r4u~o/o/?J fJ.1i- Ls_ ti/11/'f> fa,J{fJ"jz./11/'fJ

• QE-Sl.D

• EXHIBrr B REV.3

COMMONWEALTH EDISON COMPANY TABLE OF CONTENTS CALCULATION NO: HED.;.H-HSD-43 I REV 1 I PAGE 2 OF 13 SECTIONS DESCRIPTION PAGES 1 TITLE PAGE 1 2 TABLE OF CONTENTS 2 3 REVISION

SUMMARY

3 4 CALCULATION SHEET(S) 4-12 5 REVIEW CHECKLIST -*-

13 Attachments APPENDIX A A. 1-A. 3 APPENDIX B B.l QE-Sl.D EXHIBrrc REV.3

COMMONWEALTH EDISON COMPANY REVISION

SUMMARY

CALCULATION NO: - HEI>-H-HSD-43 I REV l I PAGE 3 OF 13 DESCRIPTION OF REVISIONS/REASON FOR CHANGE calculation. revised to eliminate non-QA references and inputs and to incorporate the calculation of these inputs into this document. In addition, Core Spray added to scope and a sensitivity analysis on NPSH is included.

- AFFECTED PAGES PAGES REV. DESCRIPTION l l *' . Changed Title and Equipment Nos. /System to include Core Spray 2 l Added Table of Contents 4 l Changed Purpose/Objective to include Core Spray 4,5 l Added assumptions re9arding hydraulic loss calculations and addition of Core Spray pps to scope 5 1 Removed two R. Kolflat references; added references for hydraulic loss calculations and Core Spray 6 1 Added equation for hydraulic loss calculations 7-9 l Added calculations for hydraulic losses 9 l Included discussion of NPSHR reduction due to increased temperature 10 1 Add.ed sensitivity analysis to NPSHA calculations 10 l Added Core Spray to Summary/Conclusions 11 l Added Table 2 - NPSHR values Updated Table 3 for new suction loss values 12 l Added Figure l - NPSHR reduction -vs. temperature A.1-A. 3 l New NPSH sensitivity analysis B.l l New calculation of resistance coefficient for a 2 4 x 14 reducer '

QE-51.D EXHIBrrD REV.3

Calculation No. NED-M-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA Purpose/Objective:

Calculate the Net Positive Suction Head Available (NPSHA) for the LPCI and Core Spray pumps at Dresden Station under post~

accident conditions as outlined in Reference 2, and compare with NPSH required (NPSHR) to ensure pump protection.

Assumptions/Inputs:

The NPSHA is calculated for each of the four cases analyzed by General Electric in Reference 2. Inputs to this calculation for the LPCI pumps were* taken from Tables 3, 4 and B.2 of Reference 2 and are summarized in Table 1 below:

LPCI Total Maximum Reduced Pumps Flow suppression

• Torus Case /Loop (gpm) Pool Temp(F) Pressure(psia) 3 i 10000 168 18.7 3A 2 8916 171 19.1 4 1 5000 180 19.9 4A 1 3881 186 20.6 Table 1 In addition to the assumptions made in Reference 2, the following assumptions are also made in this calculation:

1) An even split of flow is assumed between two pumps operating in.parallel; frictional losses to each pump assumed similar.

• 2) **Suction piping losses determined at 90 deg F, 5000 gpm (one pump) and 10000 gpm (two ~umps). Assumed lower temperature than Table 1 for higher kinematic viscosity and conservatively higher suction losses.

3) __strainer losses assumed to be o. 8 ft @ 5000 gpm and entrance losses assumed 0.6 ft @ 5000 gpm, 1.8 ft

@ 10000 gpm (Used Reference 11 as basis; extrapolated values provided for 5750 and 11620 gpm to 5000 and 10000 gpm respectively using quadratic relationship between

• flow and friction losses).

4) NPSHR values (Table 2) are developed based on the NPSHR curves for the LPCI and Core Spray pumps (References 5 and 6). NPSHR not reduced for higher temperatures.

5) Minimum torus level (including maximum drawdown) assumed as provided in Reference 3.

Calculation No. NED-K-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA

• 6) Assumed roughness. factor, e, for clean c9mmercial steel pipe (e = 0.00015).

7) Assumed turbulent flow through fittings.

8) Core Spray and LPCI pump suction losses similar. Also, Unit 3 LPCI/Core *spray suction losses assumed similar.

9) core Spray case bounded by LPCI case due to similar suction losses, similar NPSJ{R curves, and identical pump centerline elevations; also, Core Spray runs at a lower flow than LPCI, therefore operating at a lower NPSHR condition than LPCI.

10) Assumed all gate valves to be fully open.

References:

l) "Flow of Fluids Through Valves, Fittin9s, and Pipe",

Crane Technical Paper No. 410, 24th Printing, 1988

2) General Electric Report No. GENE-770-26-1092 "Dresden Nuclear Power Station Units 2 & 3 LPCI/Containment Cooling System Evaluation," November, 1992

3) s. Eldrid9e letter to c. Schroeder titled "Submergence of LPCI Discharge Line Post LOCA Dresden Units 2 and 3" dated September 29, 1992, chron# 0115532

4) ASME Steam Tables, 1967

5) Bingham Pump Curve No. 25355 for l2Xl4Xl4.5 CVDS, Dresden Station LPCI Pump

6) Bingham Pump curve No. 25231 for 12Xl6Xl4.5 CVDS, Dresden Station Core Spray Pump

7) Sargent & Lundy drawing M-547, LPCI pump suction

8) .Sargent & Lundy drawing M-549, Core Spray pump suction

9) "Cameron Hydraulic Data," Ingersoll-Rand Co., 16th Edition, 2nd Printing; 1984

10) "Dresden LPCI/Containment Cooling System," GE Nuclear Energy letter from s. Mintz to T. L. Chapman dated January 27, 1993

11) "Dresden Station Units 2 and 3, Quad-Cities station Units 1 and 2, NRC Docket Nos. 50-237, 50-249, 50-254, and 50-265," letter from G. J. Pliml to D. L. Ziemann dated September 27, 1976

12) "Centrifugal Pump Clinic," Karassik, Igor J., second edition, Marcel Dekker, Inc., New York, 1989

Calculation No. NED-M-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA Equations:

Suction Losses Straight piping and fitting losses are* determined using the following equation (Reference 1, page 3-4):

0.00259

• K

• Q2 hL = ( l}

d4 where: hL = frictional losses (ft)

K = resistance coefficient Q = flow (gpm) d = inner diameter of pipe (in)

The resistance coefficient, K, is the sum of the resistance coefficient for the fittings, Kf, and the resistance coefficient for the straight pipe, Kp. Kf can be obtained directly from applicable tables (Reference 9). For straight pipe, Kp is defined as:

L Kp = f (2)

D where: f = friction factor L = length of pipe (ft)

D = inner diameter of pipe (ft)

The friction factor, f, is dependent upon the pipe diameter, Reynold's number, and pipe roughness, and can be determined using the Moody dia9ram (Reference 1). Rernold's number, Re, is determined using the following equation (Reference 1, page 3-2):

50. 6

• Q *f Re = (3) d

• y where: .f = density, lb/ft3

)' = dynamic viscosity (centipoise)

i;*,.;J~

calculation No. NED-M-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSllA Evaluation - Post DBA-LOCA

.Net Positive Suction Head Net Positive Suction Head Available (NPSHA) is determined using the following equation:

NPSHA = 144

• CPt - Pvl + Z - hL (4)

J' where: Pt = Torus Pressure given in Table 1 (psia)

Pv = Vapor Pressure from Reference 4 (psia)

Z = Static Head, the minimum water elevation expected above the LPCI/Core Spray pump suction as calculated below:

Minimum Torus water level elevation 491.42' (including maximum post-LOCA draw down as discussed in Reference 3)

LPCI/CS pump suction elevation - 478.13' Static Head *13 . 2 9 '

hL = suction losses in feet Calculations:

suction Losses - One Pump The suction piping for LPCI pump 2A is shown in Reference 7 and is made up of the following components:

Line Component No. Kf a L/D Loss(ft) 2-1502-24 11 Entrance loss 90 deg elbow (LR)b

-1 ----

0.19 0.6 ID= 23.25 11 45 deg elbow 1 0.19 gate valve 1 0.10 reducing tee (~h~) 1 0.24

-- 16 1 straight pipe - 8.26 Total 0.72 8.26 0.6 2-1502A-14" reducer, 24X14 1 0.07C 90 deg elbow 2 0.78 ID= 13.25" 45 deg elbow 1 0.21 gate valve 1 0.10 strainer 4 I straight piped 1

- ---- 3.62 o.8 Total 1.16 3.62 0.8 8

from Reference 9 4 Total straight pipe length determined b from Reference 11 as the sum of all straight pipe lengths c see Appendix B minus the length of all fittings

calculation No. NED-M-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA The Reynold's number for each piping run is determined using Equation 3 (@ 90 deg F):

50.6 * (5000) * (62.116)

Re24 = = 9.0 x 10 5 (23.25) * (0.75) 50.6 * (5000) * (62.116)

= 1.6 x 10 6 (13.25) * (0.75)

The friction factor for each piping run can then*be determined using the Moody diagram for clean commercial steel pipe (Reference 1: A-25):

f24 = 0.0132 f14 = 0.0134 The resistance coefficient, K, is now be determined for each piping run utilizing Equation 2 for the straight pipe portion:

K24 = Kf + Kp

= 0.72 + (0.0132)*(8.26)

= 0.83 K14 = 1.16 + (0.0134)*(3.62)

= 1.21 Using Equation 1, the friction ioss for each piping run and

'total.suction friction losses can be determined as follows:

0.00259 x 0.83 x (5000) 2 hL24 = 0*6 I +

(23.25) 4

= 0.78 feet 0.00259 x 1.21 x (5000)2 hL14 = 0 *8 I +

(13.25) 4

= 3.34 feet hLtot = 0.78 + 3.34

= 4.12 feet @ 5000 gpm

• To determine frictional losses at any flow, the quadratic relationship between hL and Q establishes the following:

hL2 = hLl x (Q2/Ql) 2 ( 5)

0/i3 calculation No. NED-M-MSD-43 Rev 1 Dresden LPCI/Core Spray Pumps NPSHA Evaluation - Post DBA-LOCA Suction Losses - Two Pumps For two pump operation, most of the 24" line (assume all) sees full flow (10000 gpm), while each of the 14" lines that branch off of it see one-half full flow (5000 gpm). Since the 14" line was previously analyzed at 5000 gpm, only the 24" line at 10000 gpm needs to be analyzed.

The Reynold's number and friction factor for the 24" line at 10000 gpm are:

50.6 x 10000 x 62.116 Re24 = - 1. 8 x 10 6 23.25 x 0.75 f24 = 0.0125 The resistance coefficient and frictional losses for the 24" pipe at 10000 gpm are then calculated as:

K24 = Kf + Kp .

= 0.72 + (0.0125)*(8.26)

= 0.82 0.00259 x 0.82 x (10000)2 hL2 4 = 1.8' +

(23.25) 4

= 2.53 feet The suction friction losses for each pump with two pumps running .is:

hLtot = 2.53 + 3.34

= 5.87 feet @ 10000 gpm total flow NPSHA Calculations:

Using Equation 4 and the inputs provided in Table 1 and Equation 5, the NPSHA is calculated for each of the .four cases (Table 3). The required NPSH is also provided and the difference between the two is calculated. The NPSHR provided is for cold water and is not adjusted for the increased temperatures expected in the torus. This adjustment would have taken the form of a NPSHR reduction and resulted in a greater margin for NPSHA over NPSHR. From Figure 1 (Ref. 12), the reduction at 170 deg F (Cases 3 and 3A) would be about 0.3 feet, and at 180 deg F (Cases 4 and 4A) would be about 0.4 feet.

calculation No.* NED-M-MSD-43 Rev 1

• Dresden LPCZ/Core Spray Pwnps HPSHA Evaluation - Post DBA-LOCA

• The margin between available and required NPSH in Table 3 is given in feet. In order to better* understand the significance of this margin, a sensitivity analysis was performed (Appendix A) based on each of the following:

Al} torus temp.erature increase (Cases 3 and 4}

A2} torus pressure decrease (Cases 3 and 4)

AJ} ccsw initiation time increase (All cases}

In preparing this sensitivity analysis, the following conservative assumptions were made:

Al} As torus temperature increases, torus pressure remains constant.

A2} Torus temperature remains unchanged for lower torus pressures.

• AJ} Higher temperatures produced by dela¥ing the initiation of ccsw will not be accompanied by higher pressures.

Summary/Conclusions:

Post DBA-LOCA torus conditions were determined in Reference 2 and were used to calculate the available NPSH for the LPCI and Core Spray pumps at Dresden Station. The results in Table 3 indicate that the available.NPSH is greater than the required NPSH (with margin} for all four cases, and therefore adequate to protect the pumps under these conditions. While the calculations performed were for the LPCI 2A pump, the results bound the remaining LPCI pum~s as well as the Core Spray pumps for both Units based on similar suction losses, required NPSH and pump elevations.

Calculation No. Nt:.SD-43 Rev 1 Dresden LPCl/Core Spray Pumps NPSHA Evaluation ~ Post OBA LOCA Flow NPSHR Flow NPSHR (gpm) (ft) (gpm) (ft).

3500 25.0 5500 35.0 3800 25.5 5600 36.1 4000 26.0 5700 37.2 4500 27.0 5800 38.4 5000 30.0 5900 39.5 5300 33.0 6000 40.6 Table 2 Total Single Torus Torus Specific Vapor Suction Flow Pump Temp Pressure Static Volume Pressure Losses NP SHA NPSHR Margin Case (gpm) Flow (gpm) (F) (psia) Head (ftl (ft3/lb) (psia) * (ft) (ftl (ft) (ft) 3 10000 5000 168 18.7 13.29 0.01644 5.722 5.87 38.14 30.00 8.14 3A 8916 4458 171 19.1 13.29 0.016457 6.132 4.67 39.35 26.90 12.45 4 5000 5000 180 19.9 13.29 0.01651 7.511 4.12 38.62 30.00 8.62 4A 3881 3881 186 20.6 13.29 0.016547 8.568 2.48 39.48 25.70 13.78 Table 3

1::

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• Figure 1.29 NPSH reductions for pumps handling hydrocarbon liquids and high-temperature water (Courtesy Hydraulic Institute Standards Qf 1975.)

REVIEW CHECKLIST CALCULATION NO: rJt:c ... fl'\- r"\SO -'f~ I REV. { I PAGE 13 OF 13 REVIEWED BY: f)~ /(. l. I DATE: ~1.1(~~

m W2 REMARKS

,/ 0 1. IS 1ME OBJECT1VE OF 'THE ANALYSIS CLEARLY STATED7

~ 0 2. ARE ASSUMl"TIONS AND ENGINEERING JUDGEMENTS VALID AND DOaJMENTED7

~3. ARE THERE ASSUMl"TIONS 'THAT NEED VERIFICAT10N7

~ 0 4. ARE 'THE REFERENCES ILE. DRAWINGS. CODES. STANqARDSJ usTE> BY REVISION EDmON. DATE. ETC.7 r/ 0 5. IS 'THE DESIGN METHOD CORRECT AND APl'ROPRIATE FOR THIS ANALYSIS?

J 0 8. IS 'THE CALCULATION IN COMPLIANCE wmt DESIGN aurauA.

CODES. STANDARDS. AND REG. GUIDES7 d 0 7. ARE 'THE UNITS CLEARLY IDENTIFIED. AND EQUATIONS PROPERLY DERIVED AND APPUED7

., r:J 0.. . B. . ARE 'THE DESIGN INPUTS AND 'THEIR SOURCES IDEN1'1RED AND IN COMPLIANCE wmt UFSAR

• TEat SPECS7 d 0 9. ARE 1ME RESULTS COMPAT1BLE Wl1M 'THE INPU'TS AND RECOMMENDATIONS MADE1

10. INDICATE 'TYPE OF CALCULATION IHAM>-PAEPARED AND/OR COMJIV1ER.AIDEDJ AND ME'THOD OF REVIEW:

~g PBEeABEC QESI~~ ~aL~ULAilQril

'THE REVIEW OF THE HAND-PREPARED DESIGN CALCULATION WAS ACCOMPLISHED BY ONE OR A COMBINATION OF THE FOLLOWING CAS atECXEDI:

~DETAILED REVIEW OF 1ME ORIGINA{ CALCULATION z:

0 A REVIEW BY AN ALTERNATE. SJMPUFIED OR APl'ROXJMATE METHOD OF CALCULATION OF A REPRESENTATIVE SAMPLE OF REPETTT1VE CALCULATIONS OF THE cALCULATION AGAINST A SIMJLAR CALCULATION PREVIOUSLY PERFORMED 0 COMPUTER AIDED DESIGN CALCULATION m W2 m ?ill CJ 0 t 1. IS 'THE PROGRAM APPLICABLE TO ntts PROBLEM? 0 0 15. ARE THE RESULTS CONSISTENT WITH THE ASSUMl"TIONS AND THE INPUT DATA?

CJ 0 12. IS THE COMPUTER PROGRAM VALIDATED PER QP 3-547 0 0 11. IS A UST OF THE PROGRAMS USED ANO DA TC Cl CJ. 13. IS "'THE COMPUTER PROGRAM VALIDA TED BY O'THER AE'S I OF EACH COMPU~ RUN REFERENCED IN n4E ORGANIZATIONS ANO HAS IT BEEN PREVIOUSLY APf'UED TO CALaJLA TION7 NUCLEAR PRO.J£CTS7 CJ 0 17. IS THE PROGRAM VERSION AND 1rs REVISION Cl Cl 14. IS THE INPUT DATA IN CONFORMANCE WITH IDENTlFIED ON n4E COMPUTER RUN7 THE DESIGN INPUTS7 Qf*&1.0 EXHIBIT, llEV. l

I 0*1 J Calculation No. NE: SD-43 Rev 1 Dresden LPCl/Core Spray Pumps NPSHA E'.!aluatlon

• Post DBA LOCA Appendix A NPSH Margin CCSW Initiation Time Sensitivity Increase from 600 to 1800 Seconds Total Single Torus* Torus Specific Vapor Suction 1800 s 600 s Flow Pump. T~rriJ>. Pressure Static Volume Pressure Losses NPSHA NPSHR Margin Margin Case (gpm) Flow (gpm) < Jf) ..*. (psia) Head (ft) (ft3/lb) (psia) (ft) (ft) (ft) (ft) (ftl 3' 10000 5000 :112* 18.7 13.29 0.016463 6.274 5.87 36.88 30.00 6.88 8.14 3A' 8916 4458 114***

19.1 13.29 0.016474 6.566 4.67 38.35 26.90 11.45 12.45 4'

4A' 5000 3881 5000 3881 yq~~***

>188 .

19.9 20.6 13.29 13.29 0.016522 0.016559 7.851 8.947 4.12 2.48 37.84 38.60 30.00 25.70 7.84 12.90 8.62 13.78 Table A-1 *

• increased Values of Torus Temperature from Reference 10

Calculation No. NEO-M-MSD-43 Rav 1 Dresden LPCl/Cora Spray Pumps NPSHA Evaluation- Post DBA-LOCA Appendix A NPSH Margin Temperature Sensitivity Case 3: Two Pumps - 10,000 gpm - 18.7 PSIA

~-- *Ta.bl~.

/n-v-** ~

9.00 8.00

?.00 r

~ 6.00 J; 5.00 a

ID 4.00

E 3.00

2.00 en A.

z 1.00 0.00

.;,1.00I 5 170 . 175 180 185 195

-2.00 Torus Peak Temperature (F)

Case 4: One. Pump - 5000 gpm - 19.9 PSIA 9.00 8.00 7.00 6.00

.6 5.00 a

ID

E 4.00

en 3.00 A.

z 2.00.

1.00 0.00

-1.001 Torus Peak Temperature (F)

Figure A-1

Call:utatton No. NED-M-MSD-43 Rev 1 Dresden LPCl/Cora Spray Pumps NPSHA EvaJuatio~ Post DBA-LOCA Appendix A NPSH Margin Pressure Sensitivity Case 3: Two Pumps - 10,000 gpm - 186 F froff'I Tqblt.3 9.00 I 8.00 7.00

- 6.00

.6 5.00 Q

~

ID 4.00

~ 3.00

Cl) 2.00 a..

z 1.00 0.00

-1.0Q .o 19.0

-2.00 Torus Pressure (psia)

Case 4:

• One Pump - 5000 gpm - 180 F .

10.00 ff.'ore. 3 I

8.00

£ 6.00 c

1i~

ID

~

4.00

Cl) a.. 2.00 z

0.00 1 .0 *.20.0

-2.00 Torus Pressure (psia)

Figure A-2

• -. -*- ._-.. . .~ - . . .....

p°"j e. 6. I of B. I calculation Bo. BBJ>-M-HSD-43 aev 1 Dresden LPCX/Cor* Spray PUmpa 11PSBA Bvaluation - Post DBA-LOCA APPBllDll B calculation of Resistance coefficient of 24 z 14 Re4uaer From Reference 1 (A-26), the equation for the resistance coefficient of a reducer is qiven by:

K .~ .. *o.s sin (a/2) (1 - b2) (B-1)

(d2 - dl) where a a 2 tan-1 [ 2L J b -=.dl/d2 dl a small *diameter of reducer (in) d2 a larqe diameter of reducer (in)

L = lenqth of reducer (in)

For*a 24 x 14 reducer, the above parameters are defined as:

dl = 13.25 in L = assume dl + d2 d2 = 23.25 in = 36.5 in Therefore, b =0.57 and a= 15.6 deg Substitutin~ into Equation A-1, the resistance coefficient for the reducer is:

K = 0.07