W3F1-2004-0044, Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis

Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis

ML041490096
Person / Time
Site:	Waterford
Issue date:	05/26/2004
From:	Houston B Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2004-0044
Download: ML041490096 (22)

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Entergy Killona, LA 70057 Tel 504 739 6440 Fax 504 739-6698 bhousto~entergy.com Bradford Houston Director, Nuclear Safety Assurance Waterford 3 W3Fl-2004-0044 May 26, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy Letter dated April 29, 2004, "Reporting of Information under 10 CFR 50.46, Newly Identified Single Failure for Small Break LOCA Analysis of Record"

2. Entergy Letter dated April 30, 1998, 'Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis Using the ABB/CE Supplement 2 Model"

Dear Sir or Madam:

By letter (Reference 1), Entergy Operations, Inc. (Entergy) reported, pursuant to 10 CFR 50.46(a)(3)(ii), an error discovered in the Waterford Steam Electric Station, Unit 3 (Waterford 3) small break loss-of-coolant accident (SBLOCA) analysis of record (see Reference 2). That letter described the nature of the error and its effect on the current Waterford 3 emergency core cooling system (ECCS) analysis. Also in that letter Waterford 3 committed to submit the small break LOCA re-analysis results to demonstrate compliance with 10 CFR 50.46 by May 31, 2004.

The attached description of a revision to the Waterford 3 ECCS analysis of record for the SBLOCA was performed using the Supplement 2 version (referred to as the S2M or Supplement 2 Model) of the Westinghouse SBLOCA evaluation model for Combustion Engineering (CE) designed Pressurized Water Reactors (PWRs). This is the same model as used for the current analysis of record. The S2M was approved by the NRC for use by CE-designed plants on December 16, 1997.

The revised analysis utilizes the same methodology used in the analysis of record. Two of the design inputs have been changed: credit for supplemental charging flow has been eliminated and the flow curve for the high pressure safety injection pump has been revised.

As described in the attachment, the results of the revised Waterford 3 SBLOCA ECCS performance analysis conform to the ECCS acceptance criteria of 10 CFR 50.46. A table AD01

W3Fl-2004-0044 Page 2 of 3 identifying the impact of various model and input changes on the SBLOCA analysis made since the analysis of record (Reference 2) is included in the attachment.

There are no new commitments contained in this letter. If you have any questions or require additional information, please contact Jerry Burford at 601-368-5755.

Sincerely, BLH/FGB/cbh

Attachment:

Description of Analysis and Results

W3F1-2004-0044 Page 3 of 3 cc: Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector Waterford 3 P.O. Box 822 Killona, LA 70057 U.S. Nuclear Regulatory Commission Attn: Mr. Nageswaran Kalyanam MS 0-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn Attn: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. O. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 2 9 th S. Main Street West Hartford, CT 06107-2445

Attachment to W3FI -2004-0044 Description of Analysis and Results

Attachment to W3F1 -2004-0044 Page 1 of 18 Description of Analysis and Results 1.0 Introduction On March 31, 2004, Entergy Operation, Inc. (Entergy) reported the identification of a different worst case single failure for the Waterford Steam Electric Station, Unit 3 (Waterford 3) small break loss-of-coolant accident (SBLOCA) emergency core cooling system (ECCS) performance analysis (Reference 1). This worst case single failure is that of a direct current (DC) power bus. The failure of a DC power bus results in the inability to start one emergency diesel generator (EDG) and the failure of a charging loop isolation valve to remain open.

The current Waterford 3 SBLOCA analysis assumes the failure of an EDG as the worst single failure. In addition, it credits the flow from one charging pump. The charging pumps inject into two reactor coolant pump (RCP) discharge legs. Therefore, after accounting for the assumption that charging flow to the broken RCP discharge leg will not reach the reactor vessel, the SBLOCA analysis credits 50% of the flow from one charging pump reaching the reactor vessel.

With an assumed failure of a DC bus and the consequential failure of a charging loop isolation valve to remain open, one RCP discharge leg receives 100% of the charging flow. However, if that discharge leg is postulated to be the location of the break, then no charging flow is assumed to reach the reactor vessel. This is contrary to the current Waterford 3 SBLOCA analysis, which credits 50% of the flow from one charging pump. Therefore, the consequences of this worst case single failure are not bounded by the analysis and thus this single failure represents an unanalyzed condition.

This attachment describes a new SBLOCA ECCS performance analysis for Waterford 3 that does not credit any injection flow from the charging pumps. Given the fact that the new analysis does not credit injection from the charging pumps, the failure of a DC bus and the failure of an EDG are functionally equivalent with respect to their impact on the availability of ECCS equipment. Consequently, the failure of an EDG is nominally selected as the most damaging single failure assumed in the analysis.

In order to compensate for the adverse impact of the removal of credit for charging flow, the new analysis credits additional flow from the high pressure safety injection (HPSI) pump. The additional flow was obtained by removing discretionary conservatism that was included in the calculation of the HPSI pump delivery curve (i.e., HPSI pump flow versus reactor coolant system (RCS) pressure) used in the current SBLOCA analysis.

The following sections of this attachment describe the methodology, changes in plant design data, results, and conclusions of the new analysis.

2.0 Methodology The new analysis has been performed using the Supplement 2 version (referred to as the S2M or Supplement 2 Model) of the Westinghouse SBLOCA evaluation model for Combustion Engineering (CE) designed Pressurized Water Reactors (PWRs) (Reference 2). This is the same methodology used in the current Waterford 3 SBLOCA analysis, which was submitted to the NRC in Reference 12. It is also described in Sections 6.3.3.3 and 15.6.3.3.3.2 of the

Attachment to W3Fl-2004-0044 Page 2 of 18 Waterford 3 Final Safety Analysis Report (FSAR) (Reference 3). The S2M is accepted by the Nuclear Regulatory Commission (NRC) for use in CE design PWR licensing applications, including reference in plant technical specifications and core operating limits reports (Reference 4).

In the S2M evaluation model, the CEFLASH-4AS computer program (Reference 5) is used to perform the thermal hydraulic analysis of the RCS until the time the safety injection tanks (SITs) begin to inject. After injection from the SITs begins, the COMPERC-11 computer program (Reference 6) is used to perform the thermal hydraulic analysis of the RCS. The hot rod cladding temperature and maximum cladding oxidation are calculated by the STRIKIN-I1 computer program (Reference 7) during the initial period of forced convection heat transfer and by the PARCH computer program (Reference 8) during the subsequent period of pool boiling heat transfer. Core-wide cladding oxidation is conservatively represented as the rod-average cladding oxidation of the hot rod. The initial steady state fuel rod conditions used in the SBLOCA analysis are determined using the FATES3B computer program (Reference 9).

As described in Section 3.0, removal of the credit for charging flow and crediting additional HPSI pump flow results in a very small net change in safety injection flow relative to the current SBLOCA analysis. Because of this, there is very little difference in the RCS thermal hydraulic transient and the hot rod heatup transient between the new analysis and the current analysis.

Consequently, only the limiting break of the current analysis (i.e., the 0.05 ft2 /PD break in the RCP Discharge leg) was reanalyzed in the new analysis. Also, STRIKIN-I1 was not run in the new analysis since there is an insignificant difference between the current and new analyses during the forced convection portion of the transient, which lasts for approximately 300 seconds for the 0.05 ft2/PD break. Since STRIKIN-I1 was not run, the PARCH computer program was initialized with the STRIKIN-I1 results from the current analysis.

The current analysis was performed for the fuel rod conditions at the burnup that resulted in the maximum initial stored energy in the fuel. In the new analysis, additional studies were performed using PARCH to determine the fuel rod internal pressure that causes cladding rupture to occur at the time that results in the maximum peak cladding temperature (PCT) and maximum cladding oxidation.

3.0 Plant Design Data The new SBLOCA analysis uses the same plant design data used in the current analysis with the two exceptions noted in Section 1.0, namely, no credit for charging flow and the use of a revised HPSI pump delivery curve. Table 1 lists important input parameters and initial conditions used in the analysis. Except for the charging and HPSI flows, the values are the same as listed in Table 15.6-1 3a of the Waterford 3 FSAR.

Table 2 lists the HPSI pump delivery curve used in the new analysis. Figure 1 provides a comparison of the new HPSI pump delivery curve to that of the HPSI pump and charging pump flow used in the current analysis. As shown in the figure, for the RCS pressure range of interest (i.e., above approximately 500 psia), there is very little difference between the new HPSI pump delivery curve and the total safety injection (i.e., HPSI plus charging flow) credited in the current analysis.

Attachment to W3Fl-2004-0044 Page 3 of 18 The current and the new analyses were performed for up to 500 plugged tubes per steam generator (see Table 1). In order to accommodate up to 700 plugged tubes per steam generator, a PCT adder of +53 0F was previously determined. Reference 10 identified a +30F PCT adder for a minor correction to the RCP suction leg geometry. Also, Reference 11 identified a -38'F PCT adjustment for errors identified in CEFLASH-4AS.

4.0 Results Table 3 lists the peak cladding temperature and oxidation percentages that were calculated in the new analysis for the limiting break (i.e., 0.05 ft2/PD break). Times of interest are listed in Table 4. The variables listed in Table 6 are plotted as a function of time for the 0.05 ft2/PD break in Figures 2 through 9.

The results for the 0.05 ft2/PD break demonstrate conformance to the ECCS acceptance criteria as summarized below.

Parameter Criterion Results Peak Cladding Temperature <22000 F 19590 F Maximum Cladding Oxidation <17% 9.0%

Maximum Core-Wide Oxidation *1% <0.58%

Coolable Geometry Yes Yes Table 5 has been included to summarize the PCT impact of various changes since the current analysis of record. Included in this table are the effects of the two revised inputs in the new analysis.

5.0 Conclusions The results of the Waterford 3 SBLOCA ECCS performance analysis described in this attachment conform to the ECCS acceptance criteria of 10 CFR 50.46. The analysis uses the same NRC-accepted evaluation model previously used in the current analysis of record for Waterford 3. The results of the analysis are applicable to Waterford 3 for a power level of 3478 MWt (including power measurement uncertainty), a peak linear heat generation rate (PLHGR) of 13.5 kW/ft, and up to 700 plugged tubes per steam generator.

6.0 References

1. Event No. 40632, NRC Daily Events Report for April 1, 2004, "New Worst Case Single Failure May Exceed 10 CFR 50.46 Acceptance Criterion for SBLOCA.'

2. CENPD-1 37, Supplement 2-P-A, "Calculative Methods for the ABB CE Small Break LOCA Evaluation Model," April 1998.

3. Final Safety Analysis Report, Waterford Steam Electric Station, Unit No. 3, Facility Operating License No. NPF-38, Docket No. 50-382.

4. T.H. Essig (NRC) to l.C. Rickard (ABB CE), 'Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement 2, 'Calculative Methods for the C-E Small Break LOCA Evaluation Model' TAC No. M95687)," December 16,1997.

Attachment to W3F11-2004-0044 Page 4 of 18

5. CENPD-1 33P, Supplement 1, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," August 1974; Supplement 3-P, January 1977.

6. CENPD-134P, "COMPERC-1l, A Program for Emergency Refill-Reflood of the Core,"

August 1974; Supplement 1, February 1975; Supplement 2-A, June 1985.

7. CENPD-135P, "STRIKIN-Il, A Cylindrical Geometry Fuel Rod Heat Transfer Program,"

August 1974; Supplement 2, February 1975; Supplement 4-P, August 1976; Supplement 5, April 1977.

8. CENPD-138P, "PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," August 1974; Supplement 1, February 1975; Supplement 2-P, January 1977.

9. CENPD-139-P-A, UC-E Fuel Evaluation Model," July 1974; CEN-161(B)-P-A,

'Improvements to Fuel Evaluation Model," August 1989; CEN-161(B)-P, Supplement 1-P-A, 'Improvements to Fuel Evaluation Model," January 1992.

10. Entergy letter dated April 5, 2004, 'Annual Report on Westinghouse Electric Company LLC Combustion Engineering Emergency Core Cooling System Performance Evaluation Models" (W3Fl-2004-0021).

11. Entergy letter dated May 7, 2002, "Annual Report on Westinghouse Electric Company LLC Combustion Engineering Emergency Core Cooling System Performance Evaluation Models" (W3Fl-2002-0044).

12. Entergy letter dated April 30,1998, "Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis Using the ABB/CE Supplement 2 Model" (W3F1 0090).

Attachment to W3F1 -2004-0044 Page 5 of 18 Table I SBLOCA ECCS Performance Analysis Core and Plant Design Data Parameter I Value ] Units Reactor power level (including measurement uncertainty) 3478 MWt Peak linear heat generation rate (PLHGR) 13.5 kW/ft Gap conductance at the PLHGR(1 ) 1584 BTU/hr-ft 2 -OF Fuel centerline temperature at the PLHGR(') 3402 OF Fuel average temperature at the PLHGRf') 2159 OF Hot rod gas pressure(') 1113 psia Moderator temperature coefficient at initial density 0.Ox1 0 4 Ap/ F Axial shape index -0.25 asiu RCS pressure 2250 psia RCS flow rate 148x1 06 Ibm/hr 6

Core flow rate 144.15x10 Ibm/hr Cold leg temperature 557.5 OF Hot leg temperature 615.5 OF Number of plugged tubes per steam generator 500 Main steam safety valve first bank opening pressure 1117 psia Low pressurizer pressure reactor trip setpoint 1560 psia Low pressurizer pressure SIAS setpoint 1560 psia High pressure safety injection pump flow rate Table 2 gpm(psia)

Time delay for actuation of HPSI flow (with loss of offsite 30 seconds power Charging pump flow rate (to intact discharge leg) 0 gpm Safety injection tank pressure 615 psia Notes:

(1) The values for these parameters are the values for the rod average burnup of the hot rod (1000 MWD/MTU) that yields the maximum initial fuel stored energy.

Attachment to W3F1 -2004-0044 Page 6 of 18 Table 2 High Pressure Safety Injection Pump Minimum Delivered Flow to RCS (Assuming Failure of an Emergency Diesel Generator)

RCS Pressure (psia) Flow Rate (gpm)(I) (2) 0 775 92 745 231 698 352 655 486 605 609 556 798 473 901 423 978 382 1047 342 1183 249 1244 196 1287 152 1326 100 1366 0 Notes:

(1) The flow is assumed to be split equally to each of the four discharge legs.

(2) The flow to the broken discharge leg is assumed to spill out the break.

Attachment to W3F1 -2004-0044 Page 7 of 18 Table 3 SBLOCA ECCS Performance Analysis Results PeakClading Maxmum lading Maximum Core-Break Size PemeaktCadin Ma) ximudCadding( Wide Cladding Tempratre (F) xidaion(%)Oxidation (%)

0.05 ft2IPD 1959 9.0 <0.58 Table 4 SBLOCA ECCS Performance Analysis Times of Interest (seconds after break)

HPSI Flow LPSI Flow SIT Flow Break Size Reactor Trip Delivered to Delivered to Delivered to PCT Occurs a 0A RCS RCS RCS 0.05 ft2IPD 131 161 n/aM' 1740(2) 1802 Notes:

(1) Calculation completed before LPSI flow to the RCS begins.

(2) SIT injection calculated to begin but not credited in analysis.

Attachment to W3Fl-2004-0044 Page 8 of 18 Table 5 PCT Summary Table for the Waterford 3 SBLOCA ECCS Performance Analysis Temperature (OF)

Current Analysis of Record PCT 1929 Changes and Errors CEFLASH-4AS error -38 Increase in SGTP to 700 plugged tubes/SG +53 Suction leg geometry error +3 Increase in HPSI pump flow -204 Removal of credit for charging flow +216 New Analysis of Record PCT l 1959

Attachment to W3Fl-2004-0044 Page 9 of 18 Table 6 SBLOCA ECCS Performance Analysis Variables Plotted as a Function of Time Variable Normalized Total Core Power Inner Vessel Pressure Break Flow Rate Inner Vessel Inlet Flow Rate Inner Vessel Two-Phase Mixture Level Heat Transfer Coefficient at Hot Spot Coolant Temperature at Hot Spot Cladding Temperature at Hot Spot

Attachment to W3F11-2004-0044 Page 10 of 18 Figure 1 Comparison HPSI Pump Delivery Curves (Flow to the Three Intact RCP Discharge Legs) 1400 1200 1000 CO 800 Co CO CO WQ 600 a-400 200 NEW PSIPUMP LOW CURF ENT HPSI PI~MP FLOW

.----- CUR ENT HPSI PUMP FLOW CHARGING FLOW fl..

0 0 100 200 300 400 500 600 FLOW RATE, GPM

Attachment to W3FI-2004-0044 Page 11 of 18 Figure 2 0.05 ft2IPD Break in the RCP Discharge Leg Normalized Total Core Power 1.50 . . ' ,'~  ; , , -_. ,

1.25 0.0 0

~-0.75 0,00 ,.,. ... v s A e................. ....... ........ A 0

U-0.50 0

z 0.25 0 100 ZUU OUUJv A,Inn Fnn W

TIME, SEC

Attachment to W3Fl-2004-0044 Page 12 of 18 Figure 3 0.05 ft2/PD Break in the RCP Discharge Leg Inner Vessel Pressure 2000 1600 Li D 1200240 I ,,,, .....

CD 800 400 0 600 1200 1800 2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 13 of 18 Figure 4 0.05 ft2/PD Break in the RCP Discharge Leg Break Flow Rate 1200 .; , .- , .

1000 800 - . . ., .. . . . ,.

600 0

-J LiL 400 200 0 600 1200 1800 2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 14 of 18 Figure 5 0.05 ft2 1PD Break in the RCP Discharge Leg Inner Vessel Inlet Flow Rate 50000E . '

40000 30000 - _ _ _ _ _ _ _ _ _

0 1H 0000

\

0

-10000 -

0 600 1200 1800 2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 15 of 18 Figure 6 0.05 ft2IPD Break in the RCP Discharge Leg Inner Vessel Two-Phase Mixture Level 40 32 w

Uj 24 0~

16 8

0 600 1200 1800 2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 16 of 18 Figure 7 0.05 ft2/PD Break in the RCP Discharge Leg Heat Transfer Coefficient at Hot Spot 5 . . . ...

10 .. .. . .

4 10 LzL 0

e'j1 3 10

(-5 2 m- 10 10 10 10 0

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1200 1800

,,,,X II.,,.

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2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 17 of 18 Figure 8 0.05 ft2/PD Break in the RCP Discharge Leg Coolant Temperature at Hot Spot 1800 1500 1200 LL 0

uS 900 a-w Iiii 600 300 0

0 600 1200 1800 2400 3000 TIME, SEC

Attachment to W3Fl-2004-0044 Page 18 of 18 Figure 9 0.05 ft2IPD Break in the RCP Discharge Leg Cladding Temperature at Hot Spot 2200 . .i. , .,,;,.,,;,,.,....

1900 1600 0

Li 1300 1000 700 400 . ........ . , , , .,..

0 600 1200 1800 2400 3000 TIME, SEC