Text

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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Dresden Nuclear Power Station GE Nuc:ear : 1err;1 GENE-770-26-1092 DRF-T23-685 CLASS II November 1992 --*-

Units 2* and 3 LPCI/Containment Cooling System Evaluation r

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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.

- ii -

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.

- iii -

ABSTRACT

1.0 INTRODUCTION

GENE-770-26-1092 TABLE OF CONTENTS 2.0 CONTAINMENT PRESSURE ANO TEMPERATURE RESPONSE 2.1 Model Description 2.2 Analysis Assumptions 2.3 Analysis Description 2.4 Results

3.0 CONCLUSION

S

4.0 REFERENCES

APPENDICES A.

REVIEW OF NFS CALC RSA-D-92-01 B.

REDUCTION TO THE CONTAINMENT PRESSURE C.

PRIMARY SYSTEM MASS AND ENERGY RELEASE DATA FOR DRESDEN CONTAINMENT EVALUATION

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1 4

4 4

5 6

8 9

A-1 8-1 C-1

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 LPq/

Total LPCI/

Containment Containment Total Cooling*

Cooling ccsw ccsw No. of Pumps Pump Pumps Pump Case Loops**

Per Loop Flow (qpml 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 Parameter Core Thermal Power.

Vessel Dome Pressure Drywell Free (Airspace) Volume (including vent system)

Initial Suppression Chamber Free (Airspace) Volume Low Water Level (LWL)

Initial Suppression Pool Volume

,Min. Water Level Initial Drywell Pressure Initial Drywell Temperature Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace Temperature Initial Suppression Chamber Airspace Relative Humidity

.Initial Suppression Pool Temperature No. of Oowncomers Total Downcomer Flow Area In it i a 1

Oowncomer Submergence ( LWL) Units MWt psi a ft 3 ft3 ft3 psig

F psig

F ft2 ft Value Used Analysis 2578 1020 158236 120097 112000 1.25 135 20 0.15 95 100 95 96 301.6 3.67 in

GENE-770-26-1092 Table 2 - Input Parameters Used for Containment Analysis Parameter Downcomer I.D.

Vent System Flow Path Loss Coefficient (includes exit loss)

Supp. Chamber (Torus) Major Radius Supp. Chamber (Torus) Minor Radius Suppression Pool Surface Area (in contact with suppression chamber airspace)

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

- start

- full open Supp. Chamber-to-Drywell Vacuum Breaker Valve Opening Time

. Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve assembly)

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

(including exit loss)

No. of Supp. Chamber-to-Orywell Vacuum Breaker Valve Assemblies (2 valves per assembly)

LPCI/Containment Cooling Heat Exchanger Kin Containment Cooling Mode LPCI/Containment Cooling Service Water Temperature ft ft ft ft 2 psid psid sec ft2 Value Used in Analysis 2.00 5.17 54.50 15.00 9971.4 0.15 0.5 1.0 3.14 3.47 6

Btu/s-°F See Table 3

GENE-770-26~1092 Table 2 - Input Parameters Used for Containment Anaiysis Parameter LPCl/Containment Cooling Pump Heat (per pump)

Core Spray Pump Heat (per pump)

Time for Operator to turn on LPCl/Containment Cooling System in Containment Cooling mode (after LOCA signal)

Feedwater Addition (to RPV after start of event; mass and energy)

Feedwater Node **

1 2

3 4

5 Mass 1l!2ml 34658 96419 145651 91600

65072 hp hp sec Value Used in Analvsis Enthalpy

CBtu/lbml 308.0 289.2 268.7 219.8 188.4 700 800 600 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~

Maximum Suppression Pool FSAR Case No.

Temperature (*Fl 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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watev pump/HX v~~.+or Ca..se lf.
ZDFSAR/32
)R
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rn~v+ i In~rf 2..
( s~ e.1-7 C>Yl s.i.-:;.3.1)
June 1990
. 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 heat exchanger relative to the core decay heat production.
5.2.3.4 A.
Containment Capability with Respect to Metal-Water Reactions 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 heat7*
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 Parameter Core Thermal Power Vessel Dome Pressure Drywell Free (Airspace) Volume (including vent system)
Initial Suppression Chamber Free (Airspace) Volume Low Water Level (LWL)
Initial Suppression Pool Volume Min. Water Level Initial Orywel l Pressure Initial Drywel l Temperature
. Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace Temperature Initial Suppression Chamber Airspace Relative Humidity Initial Suppression Pool Temperature No. of Downcomers Total Downcomer Flow Area Initial Downcomer Submergence (LWL)
MWt psi a ft3 ft3 ft3 psig
F psig
f
. *f ft2 ft Value Used in Analysis 2578 1020 158236 120097 112000 1.25 135 20 0.15 95 100 95 96 301.6 3.67
Table 5.2;3:2 - Input Parameters Used for Containment Analysi.s Parameter Downcomer I. D.
vent System Flow Path Loss Coefficient (includes exit loss)
Supp. Chamber (Torus) Major Radius Supp. Chamber (Torus) Minor Radius Suppression Pool Surface Area (in contact with suppression chamber airspace)
Suppression Chamber-to-Drywell Vacuum Breaker Opening Diff. Press.
- start
* -
fu 11
open Supp. Chamber-to-Drywell Vacuum Breaker Valve Openini Time Supp. Chamber-to-Drywell Vacuum Breaker Flow Area (per valve assembly)
Supp. Chamber-to-Drywell Vacuum Breaker Flow Loss Coefficient (including exit loss)
No. of Supp. Chamber-to-Drywell Vacuum Breaker Valve Assemblies
* (2 va*l ves *per assembly)
LPCl/Containment Cooling Heat Exchanger K in Containment Cooling Mode LPCl/Coritainment Cooling Service Water Temperature ft ft ft ft 2 psid psid sec ft2 Value Used in Analysis 2.00 5.17 54.50 15.00 9971.4 0.15 0.5 1.0 3.14 3.47 6
Btu/s-*F See Table 5.2.3:3
Table 5.2.3:2 - Input Parameters Used for Containment Analysis Parameter LPCI/Containment Cooling Pump Heat (per pump)
Core Spray Pump Heat (per pump)
Time for Operator to turn on LPCl/Containment Cooling System in Containment Cooling mode (after LOCA signal)
Feedwater Addition (to RPV after start of event; mass and energy)
Feedwater Node **
1 2
3 4
5 Mass 1JJ2ml 34658 96419 145651 91600 65072 hp hp sec Value Used in Analysis Enthalpy *
(Btu/lbml 308.0 289.2 268.7 219.8 188.4 700 800 600 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 LPq/
LPCI/
Containment Containment Total Cooling Cooling No. of ccsw HX No. of Pumps Flow**
ccsw Pump K
Case No.
Loops*
Per Loop (gpml Pumps Flow (qpml (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)
. (gpm)
Flow (oF)
(psia)
(ft)
(ft3/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_
L ~
2
_ __L L _
20.
w a:
l (f)
(/)
w a:
Q_
I I
I I
I I
I I
I
0.
10 102 101 104 TIME - SEC figure~ - Long-ler* OBA-LOCA Orywel J. an
~.l. 1; 7 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
~
./ft
!-f\\ ! ~
l.,J *.
I w~*
~
Ge'
I
~
0 N I
-4.. ;
I,..
~
c -
1 I
(
.;.....-/
I
~
I I
c..
ID I
I
~
fD
'e I
c..
c
I
c..
I I
I"">..
.w I
i 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

*001

.-------~.
-f fTI
~
. -u
, rn
JJ D -t c
D rn
\\
60.
I I
1 Orywe l l 2 Suppression Cha ml,.!f
'10.
~
(I -
(fl Q_
\\ ~
.2
l ---
20.
w a:
J (fl
(/)
w a:.
Q_
0.
I l
l l
l I
l l
I 10 101 104 TIME - SEC figure.A' - Lon9-len1 OBA-LOCA Orywell *nd Suppre55ton CbUlber Pressure Response for C~~e 4.(
E;. 1. 3. I 0
LL el w
Cl w
a:
400.
300.
> 200.
l-a:
cc w
Q_
~
w t-
~
100.
10 I
I I
I I
\\
I I I I I
I I
102 104 THIE - SEC figure~ - long-lena DBA-LOCA Drywell Te11periture Response for Case 41
"'5,;i.1.11 105
).
LL e>
w D
w a:
300.
200.
l 100.
t-a:
a:
w Q_
~
w I-
0.
10 I
l
~.
~
I I
I I I
I l
l to2 103 104 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 Type Seals Drive Power source Speed Pump casing Impeller Shaft Code Perfonnance Characteristics -
At 0 psi reactor pressure 4 (3 required to meet design basis)
Single stage, vertical, centrifugal Mechanical Electric rnotor Nonnal auxiliary or emergency diesel 3600 rpm Cast steel Bronze Stainless steel ASME Section III B 3 pumps running Flow 5350 gpm each - 16,000 gpm total Head 263 feet
.-JiP~ow~etr~~~~E~~~~~~go~o~~h~p~each - 1800 hp total
~H !@variable) ii.~
At 200 psi. reactor pressure Flow Head dBZr (twai 1 able)
Perfonnance Characteristics -
Flow Head Power c]P~ (Available) 2675 gpm each 565 ft 490 ~each -
4"1 f 1 pump running 5990 gpm 135 ft 560 hp 40 ft).
- 8,000 gpm total 1500 hp to ta 1 Containment Cooling Service Water Pumps Number Type Po:>.,1er source Capacity Head (approximately) 4 ( 2 needed to provide required cooling capacity)
Horizontal, centrifugal Auxiliary transfonner or emergency diesel 3500 gpm each - 7000 gpm total 435 ft
TABL! 6.2.4:1
\\co(.¥\\' ~
June 1992 6.2.4-5 LlCI/CONTAINMl1fT CQOLUfG EQUIPMENT SPECI~ICAIIQMS (Contd.)
Containment Cpolin1 Scryicc Water Pymp1 Number Type Power source Capacity Bead (approximately)
Beat £xchon1era Number Heat load Primary aide flow (containment water)
Secondary aide flow (river water) dP - river water to containment water Design temperaturea River water
Containment water Primary '(shell) design presaure Secondary (tube) deaign preaaure Beat !xchon1er Code 4 (2 needed to provide required cooling capacity)
Horizontal, centrifugal Auxiliary transformer or emergency diesel 3500 gpm each - 7000 gpm total *1ee note below 435 ft each (See Section 6.2.4.5) 10,700 gpm *aee note below 7,000 gpm *aee note below 20 pd 95*r 165*F 375 pai 375 *pai
'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 Size No. nozzles (each)
Type nozzle Suppression chamber spray header Number Size No. nozzles Type 2
8 in. sch. 160
160 Fog jet 1
4 in: acb. 40 12 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.
. O.l1d ~1~ ~o~* +he. u'f\\a..l~?I?
ZFSAR92/34/48 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
'<~'((10 11a...!
1\\*~
eueheft1e* tttbes.
The 61 limit is based on the number of excess tubes 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.
u1vc?<- <
aP~
Based on the above information, the BWRED concludes that AL-6XN to replace the existing CuNi heat exchanger tubing as required "as-needed" basis) in a.ccoY-cla.ric.e. w'H-J.i
~e.. gu1del1nes d..~vc.. ~...a 111 can be used '
,~
?\\#fj J.
R~~'(erce. 7,
*6
2. 4
5
5
0. R.EW!ffCES on an J- (p ~o )(ti-\\)~
~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 If) 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, R7,
Ml5, 57,
Ml6, S'!'l
Ml9, OPl,
OPS,
( 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
Mod # UFSAR UPDATE Exhibit E 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.
N A
2.
Describe the proposed change and the reason for the change. ___.
3.
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.
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
Mod # UFSAR UPDATE Exhibit E ENC-QE-06.l Revision 5 Page 2 of 9
station/Unit
s.
6.
Exhibit E 10CFRS0.59 SAFETY EVALUATION 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.
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
Mod # UFSAR UPDATE Exhibit E ENC-QE-06.1 Revision S Page 3 of 9
Station/Unit
=D=r=e=s=d~e~n.__ _______________________________________ / ____ __
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
Mod # UFSAR UPDATE Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION Exhibit E ENC-QE-06.1 Revision 5 Page 4 of 9
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) be increased?
( ) Yes (X] No 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 important to safety increase?
( ) Yes (X] No 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 important to safety increase?
( ) Yes
[X) No 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
Mod # UFSAR UPDATE Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION Exhibit E ENC-QE-06.l Revision S Page S of 9
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
Mod # UFSAR UPDATE Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION Exhibit E ENC-QE-06.l Revision 5 Page 6 of 9
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
Mod # UFSAB UPPATE Station/Unit Exhibit E 10CFRS0.59 SAFETY EVALUATION Exhibit E ENC-QE-06.l Revision 5 Page 7 of 9
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
Mod # UFSAR UPDATE Exhibit E ENC-QE-06.l Revision 5 Page 8 of 9
Station/Unit
~D~r~e~s~d~e~n.__ _______________________________________ /~~---
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
J Mod # UFSAB UPDATE Exhibit E ENC-QE-06.l Revision 5 Page 9 of 9
Station/Unit Note:
Exhibit E 10CFRS0.59 SAFETY EVALUATION Partial Modifications and/or separate 10CFRS0.59 reviews for rtions of the work may be used to facilitate installation.
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_~-
\\ ~
1 Y. AZ (Design Superintendent/Supervisor)

7..,........... D_a_t_e __________ ___

QE-06.l DECA Version 2.0A
No.* DESIGN ISSUE E 1 E 2 E 3 E 4 E 5 E 6 E 7 E 8 Is Class 1E equipment involved?
Is there any potential for control and power circuit interaction?
Has a sneak circuit analysis been c~ieted7 Is redundancy of existing
_systems reduced or c~romised?
Are safety related circuits Isolated and separated from non-safety related circuits?
Is safety related (Class 1E) bus Integrity aiaintained7 Has diesel generator or battery loading been checked?
Are there adequate fail safe protection features for both coqxinents and systems?
Ex.
ENC*QE-06.1 Revision 5 Page 1 of 17
DESIGN ISSUES WORKSHEETS ELECTRICAL ISSUES Mod #UFSAR UPDATE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION safety related electrical or l&C system, basis described in design input doc1.111ent
separation of voltage classes, induction effects on control signals potential shorts, inadvertent connections, i.nintended operating mode backup of protection system, fire zone consideration, independent control station, interconnection of r~t system, power supply crossties buffer ~lifiers, automatic switchgear, separate cable rl61s, electrical and physical separation bus capacity, automatic Isolation, load shedding overload potential, load sequencing and shedding, i.nlnterruptible power automatic transfer, r~t systems, failure lllOde status NO NO NO NO NO NO NO NO THE CHANGE IS TO THE UFSAR ONLY, NO EQUIPMENT CHANGES ARE BEING PERFORMED THE CHANGE IS TO THE UFSAR ONLY, NO EQUIPMENT CHANGES ARE BEING PERFORMED CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING.MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES AllE BEING MADE 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 ELECTRICAL ISSUES Mod #UFSAR UPDATE Ex ENC*QE *06. 1 Revision 5 Page 2 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION E 9 E 10 E 11 E 12 E 13 E 14 E 15 Does the design provide fault trip coordination on the system and interfacing systems?
Is actuation time of protection devices and circuitry cocrpatible with all requirements?
Are in-service periodic testing and inspection of system performance addressed?
Does the modification of control panels incorporate h~n factors objectives?
(h~n factors requires a separate evaluation)
Has bypass and inoperable status indication of Class 1E protection equipment been included in the design?
Does the design adequately address Radio Frequency Interference (Rfl) and Electromagnetic Interference CEMI)?
Do system logic configuration changes alter system design?
minimize extent of outage, interaction_with load shedding, operations sequencing, timing response time, reactor trip time, containment isolation, interaction with other systems availability for testing, frequency of testing, potential for undesirable side effects control panel layout, control f~tion, separate evaluation, control room panels and remote panels verification of status, technical specification c~llance, operational requirement new off-site sources, new electrical or electronic equipnent, new on-site coanaJnication devices, hand-held radio signals logic diagram, lnstrunent loop dlagr*
NO NO NO NO NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CllAMGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CllAMGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CltAllGES ARE BEING MADE 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
DESIGN ISSUES WORKSHEETS ELECTRICAL ISSUES Mod #UFSAR UPDATE Ex.
ENC-Qf-06.1 Revision 5 Page 3 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION E 16 E 17 E 18 Are there any grounding changes or requirements?
Have Control Room Panel additions and deletions been revised for seismic qualification iq>act?
Are there any other Electrical or l&C Issues that should be addressed?
If so, list and discuss them here.
equipment ground, ground grid, disconriecting a ground equipment cha"9es, i~ct on seismic qualification of panel, panel requal ification NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE 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
No.* DESIGN ISSUE F 1 F 2 F 3 F 4 F 5 F 6 F 7 Have all ignition sources
.been adequately controlled?
Do any additional sources of energy cause the capacity to a fire zone to be exceeded?
Are all materials of construction appropriate for fire protection purposes?
Is there additional storage of contiustible material or have contiustible materials been added as part of lllOdi f !cation?
Are there any new potential paths for fire propagation or crossing of fire zone bowldaries7 Have changes coq>romlsed testing or inspection of the fire protection system?
Have any changes been lll8de that degrade required fire detection, control or protection?
.. 9 ENC*QE-06. 1 Revision 5 Page 4 of 17 DESIGN ISSUES WORKSHEETS FIRE PROTECTION ISSUES Hod #UFSAR UPDATE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION hydrogen in contairment arcing contacts, static electric charges, open flames, off-gas control contiustibles; materials that could react to produce contiustible gas, Zn or Al in contairment excessive propagation rate, controlled materials, radiation effects, potential for failure in a fire electrical insulation coatings, gas supplies, additional cable trays constitute added fire loading
holes through fire walls or stops, ducts, daq>er failure mode thermal insulation or shielding which could block access new failure lllOdes, move or penetrate fire walls, reduce capacity of water supply system, tie-In to fire detection system NO NO NO NO NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL.CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CIWIGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CIWIGES AIE BEING MADE 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
1 No.* DESIGN ISSUE F 8 Are there any other Fire Protection Issues that should be addressed? If so, list and discuss here.
KEY WORDS DESIGN ISSUES WORKSHEETS FIRE PROTECTION ISSUES Mod #UFSAR UPDATE Ex B
ENC*OE-06.1 Revision 5 Page 5 of 17 I
IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING ~E 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
No.* DESIGN ISSUE fL 1 Is there any increase in the potential for internal flooding?
Fl 2 Are any areas or equipment susceptible to flood damage?
Fl 3 Are any potential paths for flood propagation created?
Fl 4 Is the capability to isolate
.or cope with flooding reduced?
FL 5 Are there adequate design considerations to mitigate flooding?
hhiblt B ENC-QE-06.1 Revision 5 Page 6 of 17 DESIGN ISSUES WORKSHEETS FLOODING ISSUES Mod #UFSAR UPDATE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION circulating water, condenser, D~6 pipe lines, Suppression J>ool, Fan coolers, Se~vice ~ater heat exchangers, Drywell chillers*, Sprinklers, failed check valves, augunented fire protection systems Lower levels, ~atertight roocns, Electrical equipment close to floor, P~s, Motors, Air C~ressors, Electrical Buses, Breakers, direct or Indirect failure Holes through walls, floors, & doors designed to be watertight, Floor Drains, Ventilation Ducts, backflow, siphoning, site~
topography extended removal or disengagement of valves,
~
elanns, indicators, s~llng systems, opening or isolating pipeline, blocking or closing drains, s~.
leak protection or isolation devices drainage systems, barriers, separation of equipment NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO
CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE 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
DESIGN ISSUES WORKSHEETS FLOODING ISSUES Mod #UFSAR UPDATE E
B ENC*QE*06.1 Revision 5 Page 7 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION FL 6 Are there any other Flood Protection Issues that should be addressed? If so, list and discuss here.
NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE 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
DESIGN ISSUES WORKSHEETS MECHANICAL ISSUES Mod #UFSAR UPDATE EM.
ENC-QE-06. 1 Revision 5 Page 8 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION M 1 M 2 M 3 M 4 M S M 6 Are any high energy lines added or affected?
Is the vulnerability to internally generated missiles increased?
Is the vulnerability to externally generated missiles increased?.
Is there a potential for loose particles within piping systems or c~nts? If so, how is it addressed?
Could deformation or catastrophic failure ill'18ir the safety fl.Sletlon of the system, c~ts or structures being lllOdified, or other surrO\\a'lding safety related systems?
Is the safety classification of modified systems consistent with and appropriate for the safety classification of existing systems?
jet iQ'1ingement, pipe whip, special supports new missile source(s),
~
rotor breakup, valve stem ejection, pressure vessel appendages, change In missile protection requirement tornado driven object, airplane, protection for new facilities, change in missile protection requirement cleanliness requirements, heat exchanger plugging, effect on in-line devices equipment Sl4lP0rt failure results in de&radation of safety system directly or indirectly, over pressurization failure, excessive flow forces on valve stem causing 111i soperat I on lllOdif I cat I on cif Interconnecting systems, change from non-safety related to safety related at contalnnent penetration, support attachment point, c~tlblllty of appendages NO NO NO NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEiNG MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE
. BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE 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 MECHANICAL ISSUES Mod #UFSAR UPDATE EM-ENC-QE *06. 1 Revision 5 Page 9 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION M 7 M 8 M 9 M 10 Is de>\\ble valve isolation used if changes frocn class 1 to any other class or non-class portions of a 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?
Does the system have the required fail safe protection?
Is the safety f~tion of the interfacing safety systems preserved 1..f>OO failure?
Is the redundancy of existing systems reduced by Inadequate reliability?
11 there an envlronnental qualification requirement?
(envlronnental qualification requires a separate evaluation)
Are there any changes to the environnental profile of an envlrorcnental qualification zone?
contairvnent jsolation valves, safety classification change within a piping system fail open, fail close, or fail as is at both the coq>e>nent s backup system for redundancy, adequate reliability designed in for proper redwldancy certified to operate In a specified t~rature, hunldlty, and radiation envlrOn111ent; by test, by verification analysis, or a cOlllbl natl on high energy l lne routing, changes In process per-ters NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAl CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES AlE BEING MADE 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
No.* DESIGN ISSUE M 12 M 13 M 14 M 15 M 16 Is seismic qualification required?
Have all appropriate design loads (new and existing) in addition to seismic loads been identified?
Has the c~tibility of 111aterials been evaluated?
Have changes been made that could affect the NPSH for any ~?
Ara there any changes in process parameters?
.9.
ENC*Q£*06.1 Revision 5 Page 10 of 17 *
DESIGN ISSUES WORKSHEETS MECHANICAL ISSUES Mod #UFSAR UPDATE KEY W*ORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION maintain structural integrity; operate during and after seismic event; category II over category hydrodynamic loads, pipe break loads, thermal loads material considerations, prohibited materials, sealants, coatings, insulation, effect of radiation, erosion/corrosion resistance, containnent restrictions on some materials, stainless/non* stainless interfaces 111isoperation excessive pressure loss in suction piping, cavitation, fluid temperature change.
balance of flows, t~rature, pressure limitation of existing system capability, i~ct on design function NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAMGES ARE BEING MADE YES THE REANALYSIS WAS PERFORMED TO VERIFY THE ADEQUACY OF JHE LPCI SYSTEM WITH REVISED CAPABILITIES OF THE HEAT EXCHANGERS.
ALL LOADS USED WERE VERIFIED AS APPROPRIATE BEFORE COMPLETION Of THE ANALYSIS.
NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAMGES ARE BEING MADE YES THE ANALYSIS PRCX>UCED POTENTIAL INCREASED TEMPERATURES FOR THE TORUS WATER WHICH IS THE SUCTION SCJURCE FOR THE LPCI Pt.tlPS.
THE REQUIRED AMO ACTUAL NPSHs FOR THE PUMPS HAS BEEN CALCULATED TO VERIFY ACCEPTABILITY.
YES THE ANALYSIS USED CHANGED PARAMETERS FOR THE HEAT REMOVAL CAPAlllLITIES Of THE LPCI HEAT EXCHANGERS BASED ON REDUCED FLOWS THROUGH THE IOI FROM BOTH LPCI AND CCSW.
THESE PARAMETERS WERE INDEPENDENTLY VERIFIED PRIOR TO USE IN THE ANALYSIS.
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
DESIGN ISSUES WORKSHEETS MECHANICAL ISSUES Mod #UFSAR UPDATE Ex.
ENC*OE *06. 1 Revision 5 Page 11 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION M 17 M 18 M 19
" 20 Valve Performance as it relates to system function:
can the valve be placed and maintained in the appropriate position for normal system operation, abnormal system operation, and testing mode?
valve, containnent isolation valves, valve orientation/configuration, Design Basis Event, valve closure time; isolation logic changes 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)?
Have short*term and long*term containment isolation containment isolation requirements been satisfied?
Have the rules for single failure criteria single failure criteria been applied correctly?
Are there any other Mechanical Issues that should be addressed? If so, list and discuss here.
NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE BEING MADE YES THE ANALYSIS USES THE LIMITING CASE OF PlMP AVAILABILITY BASED ON A LOCA/LOOP SCENARIO.
NO NONE 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
DESIGN ISSUES WORKSHEETS OPERATIONAL ISSUES.
Mod #UFSAR UPDATE Exti B
ENC*OE-06.1 Revision 5 Page 12 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION OP 1 OP 2 OP 3 OP 4 OP 5 OP 6 Will the ope rat.i ng conditions of this or any other system be changed?
Will the operation of any other system have any effect on the system being modified?
Will the change have any iq>act on adjacent systems?
Can the change affect the operation of another system indirectly?
Has the i~ct on operability tests been considered?
Are there any other Operational Interaction Issues that should be addressed? If so, list and discuss them here.
teq>erature, pressure, flow, cooling water supply, electrical power interr~tions shared source of power system fluid, interlocks, emergency power priorities failure modes, reduction in availability or reliability shared systems, cascading effect, ripple effect surveillance, operability test, channel check,
cal lbratlon YES THE OPERATING PARAMETERS AND TECHNICAL SPECIFICATION VALUES FOR THE LPCI AND CCSW PUMP SURVEILLANCES HAVE BEEN VALIDATED BY THIS ANALYSIS.
NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PH.YSICAL CHANGES ARE BEING MADE NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE
.NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE YES THE TECH SPEC SURVEILLANCE VALUES FOR LPCI AND CCSW Pt.ICP PERFORMANCE HAVE BEEN VALIDATED BY THIS ANALYSIS.
NO NONE 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.
DECA. Version 2.0A OE-06.1
DESIGN ISSUES WORKSHEETS RADIOLOGICAL ISSUES_
Hod #UFSAR UPDATE Exli B
ENC*QE-06.1 Revision 5 Page 13 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION R 1 R 2 R 3 R 4 R 5 R 6 R 7 Are there any changes that affect the engineered safety feature ventilation system?
Are there any changes to the controlled leakage systems CBWR), such as a change in back pressure?
Are stores of persOl'Vlel protective equipment preserved?
Are there any effects on radiation detection and monitoring or alarm systems?
Are there any effects on containnent isolation systems, ventilation systems or contairwient cleanup system?
Has separation or primary/secondary coolant systems (PWR> or contalnnent drywell (BWR) been maintained?
Are there any effects on fission product control for Incidents/accident or post accident cleanup and monitor points?
wet HEPA filters, cross*
connection, bypass or leakage high filter pressure drop, backup through air Intakes, strilctural integrity emergency air supplies for control roocn personnel, emergency breathing air suppt'ies, i~ired access false readings due to placement, unintended shielding, side effects of enclosures reliability, operability, access, containnent spra_y system, iodine removal secondary side detection system, equlpnent leakage, bcK.rdary changes contalnnent spray
cleanup system NO NO NO NO NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CKAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE 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
No.* DESIGN ISSUE R 8 R 9 R 10 Have *adequate provisions been made to ~ontrol effluent containnent levels?
Is there any potential for additional radiation exposure?
Are there any other Radiological Issues that should be addressed? If so, list and discuss here.
hhl 8
ENC*QE-06. 1 Revision 5 Page 14 of 17 DESIGN ISSUES WORKSHEETS RADIOLOGICAL ISSUES*
Hod #UFSAR UPDATE KEY WORDS monitoring required, human error.protection, potential releases, s~
cont Biii nation decontamination, ALARA, recb:tion in shielding IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHAllGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL *CHAllGES ARE BEING MADE NONE 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
I DESIGN ISSUES WORKSHEETS SITE RELATED ISSUES Mod #UFSAR UPDATE E;x ENC-QE-06. 1 Revision 5 Page 15 of 17 I
Ho.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION s 1 s 2 s 3 s 4 s 5 s 6 s 7 Is there any change in the exclusion area or site boundary conditions which would increase the on-site or off-site dose rates?
Is the site radioactive material inventory control affected?
Are release and dispersion of effluents affected?
Are there any changes affecting protection of safety class structures from natural phenomena and meteorological conditions (tornados, rain loads, snow loads)?
Are there any potential effects on security barriers or controlled access?
Are any potential hazards added to the site or exclusion area?
Are there any changes to cool! ng water S'-'3Pl y capacity or characteristics?
Change the fence line, construct a new building containing ~adioactive materials, relocate activated materials.
quantity or c~sition of radioactive materials on site - increased or changed stack height change, concentration of radwaste, or other factors affecting effluent pathways, contairment isolation valve leak rates or closure times failure effects of non-safety related structure or syste11, change to surface water control structures, secondary effects placing equipment In close proximity to guardhouse or security equipment fire source, explosive material, toxic material, radwaste material, on-site or off-site, permanent or teq>e>rary.
quantity, t~rature, aedi..nt content, aquatic growth potential, flowrates,
~
curve changes, etc.
NO NO NO NO NO NO YES CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE
~
CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE THE REVISED PARAMETERS USED IN THE ANALYSIS ACaJMMll>ATE THE REDUCED CAPABILITY OF THE LPCI HEAT EXCHANGERS FOR HEAT IEJIJVAL AT REDUCED CCSW Flo.IS.
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
DESIGN ISSUES WORKSHEET~
SITE RELATED ISSUES Mod #UPSAR UPDATE Exli B
ENC*Q£*06.1 Revision 5 Page 16 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS POR CONCLUSION s 8 s 9 s 10
. s 11 ls the stability of subsurfac~
materials or foundations for Class 1 structures affec~ed directly or indirectly?
Is plant access altered or affected?
Yill site topography changes increase the potential for external flooding?
Are there any other Site Related Issues that should be addressed? If so, list and discuss here.
ground water level, soil ph, soil response to excitation, excavating near existing structures, subsidence roadway or railroad changes, GSEP, access gate change, underground tl.l'INll excavation, topography NO NO NO NO CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NONE 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
'°"1 DESIGN ISSUES WORKSHEETS STRUCTURAL ISSUES -
Hod #UFSAR UPDATE Revision 5 Page 17 of 17 No.* DESIGN ISSUE KEY WORDS IS ISSUE RELEVANT?
PROVIDE BASIS FOR CONCLUSION ST 1 ST 2 ST 3 ST 4 ST 5 ST 6 ST 7 What is the seismic classification of the structure?
Is the response characteristic of the existing structure changed by the modification?
Does the modification degrade the structure integrity of the existing structure?
Does the modification create the possibility of failure
due to failure of non-seismic equipment affecting nearby seismic category I equipment?
Are there any changes that would affect testing ard/or in-service inspection of the structure?
Has qualification by testing, as opposed to analysis, been considered for seismic structures or COlllpOnef'lts?
Are there any other structural issues that should be eddressed? If so, list and discuss here.
Category I or non-seismic subsystem analysis, fln:tamental frequenc.,*,
stiffness, coupling, adding or redistributing mass enlarge openings, create nunerous discontinuities, additional loads, penetrations, cuaulative effects Seismic II over I, non-seiSllic/non*safety structures or equipment obstruct surface, reduce availability for testing, restrict access purchase of seismically qualified structures or c~ts, size limit, weight ll*lt YES NO NO NO NO NO 110 THE CONTAINMENT STRUCTURE IS SEISMIC CATEGORY I.
THE LOCA ANALYSIS Oii LONG TERM SUPPRESSION POOL HEAT UP IS NOT DEPENDENT Oii THE SEISMIC QUALIFICATION OF THE STRUCTURE.
CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMEllT/PLANT PHYSICAL CHANGES ARE BEING MADE CHANGES ARE TO THE UFSAR ONLY. NO EQUIPMENT/PLANT PHYSICAL CHANGES ARE BEING MADE NONE 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
To:
C. Schroeder
Subject:
Post DBA-LOCA LPCI NPSHA Evaluation November 30, 1992 Cl1ROIJ # l'i'f no
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.
cc:
S. Eldridge R. Kolflat prepared by:_* ~-i---.
..... _P_al_as __ *-----
approved
Calculation No. NED-M-MSD-43 Dresden LPCI Pumps NPSHA Evaluation-Post DBA-LOCA lotal Single Flow Pump Torus Case (gpm) Flow (gpm) Temp (F) 3 10000 5000 168 3A 8916 4458 171 4
5000 5000 180 4A 3881 3881 186
(
93031~~ ~~gg~;-j ' :\\LJ. '.
\\
~DR..
... Ptm..
)
Torus Specific Vapor Suction Pressure Static Volume Pressure Piping (psia)
Head (ft)
(ft3/lb)
(psia)
Losses (ft) 18.7 13.32 0.01644 5.7223 4.72 19.1 '
13.32 0.016457 6.1318 3.75 19.9 13.32 0.01651 7.511 3.77 20.6 13.32 0.016547 8.568 2.27 TABLE I NPSHA NPSHR Margin (ft)
(ft)
(ft) 39.32 30.00 9.32 40.30 26.90 13.40 39.00 30.00 9.00 39.72 25.70 14.02
Ql41.0 UHlllTI COMMONWEALTH EDISON COMPANY TITLE PAGE.
CALCULATION NO.
~ED-M - MSC - 'f'3
. PAGE 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)
~(:,) - t5o~ A/e/tjD STATION/UNIT Df!rJ.en ~ g.3 SYSTEM l-/'CJ REV.
CHRON #
PREPARER
DATE REVIEWER DATE APPROVER DATE 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
=
Torus Static (ft)
Pressure
+
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 (includin9 maximum post-LOCA draw down as discussed in Reference 3)
LPCI pump suction elevation Static Head 491. 5' 478.13' 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 using suction losses from References 1 and 5. (Tables 2 and 3)
LPCJ N'PSHA calculations; 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 m
tQ
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0
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1. IS ntE oeJECTM OF TME ANALYSIS a.EARLY STA1ED7
2. ARE ASSUIWT10NS AlllJ ENOMEUNG JUDGEMENTS VAUD AND DOCUMENTED?
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 I. IS ntE DUICJN MITHOO CGRMCT AND AP...a....,.TE FOR TlfJS ANALYSIS?
I. IS ntE CALCU~110N W CDMPUANCE wmt DUJON a111A1A.
C:OOU. STANDAADI. AND RICI. CIUIJll7
7. NIE 1HI UNJ18a.IAALY1811'F1ED. AND EQUATIONS PROPERLY DERNID AND Af'll\\B7 I. ARE 'TME DESIGN INPUTI AND THiii' SOURCES l>IHn:a AND W COMPUANCE Wint URAR
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~~.) '----------------------------------------------------------------------------------
QE-Sl.D
EXHIBrr B REV.3 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.,_
r4u~o/o/?J fJ.1i-Ls_ ti/11/'f> fa,J{fJ"jz./11/'fJ 1 198391
QE-Sl.D EXHIBrrc REV.3 COMMONWEALTH EDISON COMPANY TABLE OF CONTENTS CALCULATION NO:
HED.;.H-HSD-43 I REV SECTIONS DESCRIPTION 1
TITLE PAGE 2
TABLE OF CONTENTS 3
REVISION
SUMMARY
4 CALCULATION SHEET(S) 5 REVIEW CHECKLIST Attachments APPENDIX A APPENDIX B 1 I PAGE 2
OF 13 PAGES 1
2 3
4-12 13 A. 1-A. 3 B.l
QE-51.D EXHIBrrD REV.3 COMMONWEALTH EDISON COMPANY REVISION
SUMMARY
CALCULATION NO: -
HEI>-H-HSD-43 I REV l I PAGE 3
OF 13 PAGES l
2 4
4,5 5
6 7-9 9
10 10 11 12 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.
REV.
l l
l l
1 1
l l
1 l
l l
AFFECTED PAGES DESCRIPTION
'. Changed Title and Equipment Nos. /System to include Core Spray Added Table of Contents Changed Purpose/Objective to include Core Spray Added assumptions re9arding hydraulic loss calculations and addition of Core Spray pps to scope Removed two R. Kolflat references; added references for hydraulic loss calculations and Core Spray Added equation for hydraulic loss calculations Added calculations for hydraulic losses Included discussion of NPSHR reduction due to increased temperature Add.ed sensitivity analysis to NPSHA calculations Added Core Spray to Summary/Conclusions Added Table 2 - NPSHR values Updated Table 3 for new suction loss values A.1-A. 3 l
Added Figure l - NPSHR reduction -vs. temperature New NPSH sensitivity analysis B.l l
New calculation of resistance coefficient for a 2 4 x 14 reducer
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 =
d
y where:.f = density, lb/ft3
)' = dynamic viscosity (centipoise)
(3)
calculation No. NED-M-MSD-43 Rev 1 i;*, ~
.;J 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 0.6 90 deg elbow (LR)b 1
0.19 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 1
o.8 4 I straight piped 3.62 Total 1.16 3.62 0.8 8 from Reference 9 b from Reference 11 c see Appendix B 4 Total straight pipe length determined as the sum of all straight pipe lengths 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 Equation Reynold's number for each piping 3 (@ 90 deg F):
Re24 =
50.6 * (5000) * (62.116)
(23.25) * (0.75) 50.6 * (5000) * (62.116)
(13.25) * (0.75) run is determined using
= 9.0 x 105
= 1.6 x 106 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)
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.
0/i3 The Reynold's number and friction factor for the 24" line at 10000 gpm are:
Re24 =
50.6 x 10000 x 62.116 23.25 x 0.75
1. 8 x 106 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 hL24
= 1.8' +
0.00259 x 0.82 x (10000)2 (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}
A2}
AJ}
As torus temperature increases, torus pressure remains constant.
Torus temperature remains unchanged for lower torus pressures.
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
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EXHIBIT, llEV. l 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'ROPRIA TE 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 0
A REVIEW BY AN ALTERNATE. SJMPUFIED OR APl'ROXJMA TE METHOD OF CALCULATION z:
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
Calculation No. NE:
SD-43 Rev 1 I
0*1 J 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'
5000 5000 yq~~***
19.9 13.29 0.016522 7.851 4.12 37.84 30.00 7.84 8.62 4A' 3881 3881
>188..
20.6 13.29 0.016559 8.947 2.48 38.60 25.70 12.90 13.78 Table A-1 *
increased Values of Torus Temperature from Reference 10
~
J; a..
ID
E
en A. z
.6 a..
ID
E :
en A. z Calculation No. NEO-M-MSD-43 Rav 1 Dresden LPCl/Cora Spray Pumps NPSHA Evaluation-Post DBA-LOCA 9.00 8.00
?.00 6.00 5.00 4.00 3.00 2.00 1.00 0.00
.;,1.00I
-2.00 9.00 8.00 7.00 6.00 5.00 4.00 3.00 2.00.
1.00 0.00
-1.001 5
Appendix A NPSH Margin Temperature Sensitivity Case 3: Two Pumps - 10,000 gpm - 18.7 PSIA
~-- *Ta.bl~ ~
/n-v-**
r 170.
175 180 185 Torus Peak Temperature (F)
Case 4: One. Pump - 5000 gpm - 19.9 PSIA Torus Peak Temperature (F)
Figure A-1 195
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
4.00
~
ID
~ 3.00 2.00 Cl) a..
1.00 z
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) 2.00 a.. 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) tan-1 [
(d2 - dl) J where a a 2 2L 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 d2 = 23.25 in Therefore, b = 0.57 and L = assume dl + d2
= 36.5 in a= 15.6 deg Substitutin~ into Equation A-1, the resistance coefficient for the reducer is:
K = 0.07

ML17252A753

Text

1.0 INTRODUCTION

3.0 CONCLUSION

4.0 REFERENCES

References:

Reference:

Subject:

References:

SUMMARY

SUMMARY

References:

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