CPSES-200701238, Submittal of the CPSES Units 1 and 2 Large and Small Break LOCA Analyses

Submittal of the CPSES Units 1 and 2 Large and Small Break LOCA Analyses

ML072200641
Person / Time
Site:	Comanche Peak
Issue date:	07/31/2007
From:	Blevins M, Madden F Luminant Power, TXU Generation Co, LP
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CPSES-200701238, TXX-07107
Download: ML072200641 (107)

Text

Mike Blevins Luminant Power Senior Vice Piesident P 0 Box 1002 O

Chief Nuclear Officer 6322 North FM 56 mike-blevins@luminanLcom Glen Rose, TX 76043 Lum inant 2548975209 C 81t7 559 9085 F 254 897 6652 CPSES-200701238 Ref. #

10CFR50.46 Log # TXX-07107 July 31, 2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 SUBMITTAL OF THE CPSES UNITS 1 AND 2 LARGE AND SMALL BREAK LOCA ANALYSES

REFERENCE:

1. Letter logged TXX-07063, dated April 4, 2007 from Mike Blevins to the NRC.

2. Letter logged TXX-07047, dated Feb, 22, 2007 from Mike Blevins to the NRC.

Dear Sir or Madam:

Per Reference 1, TXU Generation Company LP, (Luminant Power) submitted proposed changes to the Comanche Peak Steam Electric Station (CPSES) Unit 1 and Unit 2 Technical Specifications to allow the use of several NRC-approved accident analysis methodologies to be used to establish core operating limits. Included in that submittal were different m thods for analyzing the small break loss of coolant accident (LOCA) and the large break LOCA. As prescribed in the NRC's Safety Evaluations approving the generic use of these methods, and in compliancl with 10CFR50.46, Luminant Power is herein transmitting the evaluation models and results dev loped in accordance with those methodologies for NRC review.

In addition, per Reference 2, Luminant Power comnlitted to submit to the NRC, for review, a plant-specific Westinghouse NOTRUMP-based small break LOCA evaluation model for CPSES Unit 1 for application to the beginning of Cycle 14 operation in the Fall of 2008. (Commitment Number 27436).

The evaluation models transmitted herein satisfy that commitment.

The evaluation models and results for the large breaý LOCA analyses for Comanche Peak Units 1 and 2, performed using the ASTRUM Best-Estimate LOCA ýmethodology, as described in Reference 1, are included as Attachments 1 and 2, respectively. The Comanche Peak Units 1 and 2 evaluation models for the NOTRUMP-based small break LOCA analyses are presented in Attachment 3.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance

,1 Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

U. S. Nuclear Regulatory Commission TXX-07107 Page 2 07/31/2007 If you have any questions regarding this submittal, please contact Mr. J. D. Seawright at (254) 897-0140.

Sincerely, TXU Generation Company LP By:

TXU Generation Management Company LLC, Its General Partner Mike Blevins By: J4u 2(72...

Fred W. Madden Director, Oversight & Regulatory Affairs Attachments - 1. APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 1

2. APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 2

3. APPLICATION OF WESTINGHOUSE SMALL BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNITS 1 AND 2 c -

B. S. Mallett, Region IV B. K. Singal, NRR Resident Inspectors, CPSES to TXX-07107 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 1 to TXX-07107 Page 1 of 28 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 1

Background

A best-estimate loss of coolant accident analysis has been completed for Comanche Peak Unit 1. This license amendment request (LAR) for operating license NPF-87 (Comanche Peak Unit 1) requests approval to apply the Westinghouse best-estimate large break loss of coolant accident (LOCA) analysis methodology.

Westinghouse obtained generic NRC approval of its original topical report describing best-estimate large break LOCA methodology in 1996. NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report [1]. This methodology was later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report [2].

Westinghouse recently completed a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 9 5 th percentile. This method is still based on the Code Qualification Document (CQD) methodology [1],[2] and follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics.

The Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A [3].

This report summarizes the application of the Westinghouse ASTRUM BELOCA evaluation model to Comanche Peak Unit 1 for the large break LOCA accident analysis. Table 1 lists the major plant parameter assumptions used in the BELOCA analysis for Comanche Peak Unit 1.

Both Luminant Power and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with vendor internal processes for assessing evaluation model changes and errors, are used to identify the need for LOCA analyses impact assessments.

Method of Thermal Analysis for Comanche Peak Unit 1 When the Final Acceptance Criteria (FAC) governing the loss-of-coolant accident (LOCA) for Light Water Reactors was issued in 10 CFR 50.46 [4], both the Nuclear Regulatory Commission (NRC) and the industry recognized that stipulations of Appendix K were highly conservative.

That is, using the then accepted analysis methods, the performance of the Emergency Core Cooling System (ECCS) would be conservatively underestimated, resulting in predicted peak clad temperatures (PCTs) much higher than expected. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the NRC began a large-scale confirmatory research program with the following objectives:

to TXX-07107 Page 2 of 28

1.

Identify, through separate effects and integral effects experiments, the degree of conservatism in those models required in the Appendix K rule. In this fashion, those areas in which a purposely prescriptive approach was used in the Appendix K rule could be quantified with additional data so that a less prescriptive future approach might be allowed.

2.

Develop improved thermal-hydraulic computer codes and models so that more accurate and realistic accident analysis calculations could be performed. The purpose of this research was to develop an accurate predictive capability so that the uncertainties in the ECCS performance and the degree of conservatism with respect to the Appendix K limits could be quantified.

Since that time, the NRC and the nuclear industry have sponsored reactor safety research programs directed at meeting the above two objectives. The overall results have quantified the conservatism in the Appendix K rule for LOCA analyses and confirmed that some relaxation of the rule can be made without a loss in safety to the public. It was also found that some plants were being restricted in operating flexibility by overly conservative Appendix K requirements. In recognition of the Appendix K conservatism that was being quantified by the research programs, the NRC adopted an interim approach for evaluation methods. This interim approach is described in SECY-83-472 [5]. The SECY-83-472 approach retained those features of Appendix K that were legal requirements, but permitted applicants to use best-estimate thermal-hydraulic models in their ECCS evaluation model. Thus, SECY-83-472 represented an important step in basing licensing decisions on realistic calculations, as opposed to those calculations prescribed by Appendix K.

In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, "ECCS Evaluation Models", to permit the use of a realistic evaluation model to analyze the performance of the ECCS during a hypothetical LOCA. This decision was based on an improved understanding of LOCA thermal-hydraulic phenomena gained by extensive research programs.

Under the amended rules, best estimate thermal-hydraulic models may be used in place of models with Appendix K features. The rule change also requires, as part of the LOCA analysis, an assessment of the uncertainty of the best estimate calculations. It further requires that this analysis uncertainty be included when comparing the results of the calculations to the prescribed acceptance criteria of 10 CFR 50.46. Further guidance for the use of best estimate codes is provided in Regulatory Guide 1.157 [6].

To demonstrate use of the revised ECCS rule, the NRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology [7].

This method outlined an approach for defining and qualifying a best estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis.

A LOCA evaluation methodology for three and four loop pressurized water reactor (PWR) plants based on the revised 10 CFR 50.46 rules was developed by Westinghouse with the support of EPRI and Consolidated Edison and has been approved by the NRC [1]. This methodology was to TXX-07107 Page 3 of 28 later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report [2].

More recently, Westinghouse developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical TReatment of Uncertainty Method [3]. This method is still based on the CQD methodology and follows the steps in the CSAU methodology.

However, the uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing calculations in WCAP-1 6009-P-A [3]. The ASTRUM methodology remains applicable to three and four loop PWRs, as well as 2-loop Westinghouse plants with UPI. This methodology was also extended to CE design PWRs.

The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95th percentile of the Peak Clad Temperature (PCT), Local Maximum Oxidation (LMO), and Core Wide Oxidation (CWO) with 95% confidence level. These parameters are needed to satisfy the 10 CFR 50.46 criteria with regard to PCT, LMO, and CWO.

Downcomer boiling is considered as appropriate in the ASTRUM methodology. The WCOBRA/TRAC computer code determines if downcomer boiling will occur for a particular transient. If downcomer boiling is determined to occur in a transient, WCOBRA/TRAC includes the effects of downcomer boiling in the transient calculation.

This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A [3] as applicable to the ASTRUM methodology. Section 13-3 of WCAP-16009-P-A [3] was found to acceptably disposition each of the identified conditions and limitations related to WCOBRA/TRAC and the CQD uncertainty approach per Section 4.0 of the ASTRUM Final Safety Evaluation Report appended to this WCAP. The Best Estimate LBLOCA analysis and associated model for Comanche Peak Unit 1 is unit-specific.

Description of a Large Break LOCA Transient Before the break occurs, the RCS (Reactor Coolant System) is assumed to be operating normally at full power in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. A large break is assumed to open instantaneously in one of the main RCS cold leg pipes.

Immediately following the cold leg break, a rapid system depressurization occurs along with a core flow reversal due to a high discharge of sub-cooled fluid into the broken cold leg and out of the break. The fuel rods go through departure from nucleate boiling (DNB) and the cladding rapidly heats up, while the core power decreases due to voiding in the core. The hot water in the core, upper plenum, and upper head flashes to steam, and subsequently the cooler water in the lower plenum and downcomer begins to flash. Once the system has depressurized to the accumulator pressure, the accumulator begins to inject cold borated water into the intact cold to TXX-07107 Page 4 of 28 legs. During the blowdown period, a portion of the injected Emergency Core Cooling System (ECCS) water is calculated to be bypassed around the downcomer and out of the break. The bypass period ends as the system pressure continues to decrease and approaches the containment pressure, resulting in reduced break flow and consequently, reduced core flow.

As the refill period begins, the core continues to heat up as the vessel begins to fill with ECCS water. This phase continues until the lower plenum is filled, the bottom of the core begins to reflood, and entrainment begins.

During the reflood period, the core flow is oscillatory as ECCS water periodically rewets and quenches the hot fuel cladding, which generates steam and causes system re-pressurization.

The steam and entrained water must pass through the vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out of the break. This flow path resistance is overcome by the downcomer water elevation head, which provides the gravity driven reflood force. The pumped upper plenum and cold leg injection ECCS water aids in the filling of the vessel and downcomer, which subsequently supplies water to maintain the core and downcomer water levels and complete the reflood period.

ASTRUM Analysis Results for Comanche Peak Unit 1 The results of the Comanche Peak Unit 1 ASTRUM analysis are summarized in Table 2.

Table 3 contains a sequence of events for the limiting PCT transient.

The scatter plot presented in Figure 1 shows the effect of the effective break area on the analysis PCT. The effective break area is calculated by multiplying the discharge coefficient CD with the sample value of the break area, normalized to the cold-leg cross sectional area.

Figure 1 is provided to show the break area is a significant contributor to the variation in PCT.

From the 124 calculations performed as part of the ASTRUM analysis, different cases proved to be the limiting PCT and limiting LMO transient for Comanche Peak Unit 1. Figure 2 shows the predicted clad temperature transient at the PCT limiting elevation for the limiting PCT case.

Figure 3 presents the clad temperature transient predicted at the LMO elevation for the limiting LMO case. Due the low PCT results, CWO remained on the order of 0 percent for all cases.

Figure 4 through 17 illustrate the key major response parameters for the limiting PCT transient.

The reference point for the lower plenum liquid level presented in Figure 11 is the bottom of the vessel (10.1 feet below the bottom of the active fuel). The reference point for the downcomer liquid level presented in Figure 12 is the point at which the outside of the core barrel, if extended downward, intersects with the vessel wall (6.1 feet below the bottom of the active fuel). The reference point for the core collapsed liquid levels presented in Figures 13 and 16 is the bottom of the active fuel.

The containment backpressure utilized for the LBLOCA analysis compared to the calculated containment backpressure is provided in Figure 18. The worst single failure for the LBLOCA analysis is the loss of one train of ECCS injection (consistent with the ASTRUM Topical);

to TXX-07107 Page 5 of 28 however, all containment systems which would reduce containment pressure are modeled for the LBLOCA containment backpressure calculation.

10 CFR 50.46 Requirements It must be demonstrated that there is a high level of probability that the following limits set forth in 10 CFR 50.46 are met:

(b)(1)

The limiting PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT for the limiting case is 1492°F for Comanche Peak Unit 1, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., "Peak Clad Temperature less than 2200°F", is demonstrated. The result is shown in Table 2.

(b)(2) The maximum cladding oxidation corresponds to a bounding estimate of the 95th percentile LMO at the 95-percent confidence level. Since the resulting LMO for the limiting case is 0.23 percent for Comanche Peak Unit 1, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Local Maximum Oxidation of the cladding less than 17 percent", is demonstrated. The result is shown in Table 2.

(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 9 5 th percentile CWO at the 95-percent confidence level. While the limiting LMO is determined based on the single Hot Rod, the CWO value can be conservatively chosen as that calculated for the limiting Hot Assembly Rod (HAR) when there is significant margin to the regulatory limit. The limiting HAR total maximum oxidation is 0 percent for Comanche Peak Unit 1. Thus, a detailed CWO calculation is not needed because the calculations would include many lower power assemblies and the outcome would always be less than the limiting HAR total maximum oxidation. Therefore, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation less than 1 percent", is demonstrated. The result is shown in Table 2.

(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The grid crush calculations currently in place for Comanche Peak Unit 1 remain unchanged with the application of the ASTRUM methodology [3], therefore, acceptance criterion (b)(4) is satisfied.

to TXX-07107 Page 6 of 28 (b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS. Long-term cooling is dependent on the demonstration of continued delivery of cooling water to the core. The actions, automatic or manual, that are currently in place at Comanche Peak Unit 1 to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology [3].

Based on the ASTRUM analysis results (see Table 2), it is concluded that Comanche Peak Unit 1 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

to TXX-07107 Page 7 of 28 References

[1]

Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).

[2]

Dederer, S. I., et. al., 1999, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A, Revision 1 and WCAP-14450, Revision 1 (Non-Proprietary).

[3]

Nissley, M. E., et.al., 2005, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

WCAP-16009-P-A and WCAP-16009-NP-A (Non-Proprietary).

[4]

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 4, 1974.

[5]

Information Report from W.J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods," SECY-83-472, November 17, 1983.

[6]

"Best Estimate Calculations of Emergency Core Cooling System Performance,"

Regulatory Guide 1.157, USNRC, May 1989.

[7]

Boyack, B., et. al., 1989, "Quantifying Reactor Safety Margins: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-of-Coolant-Accident," NUREG/CR-5249.

to TXX-07107 Page 8 of 28 Table 1 - Major Plant Parameter Assumptions Used in the BELOCA Analysis for Comanche Peak Unit 1 Parameter Value Plant Physical Description SG Tube Plugging

< 10%

Plant Initial Operating Conditions Reactor Power

< 3612 MWt (+/- 0.6% uncertainties)

" Peaking Factors Fa - 2.5 FAHN-1.60 Axial Power Distribution See Figure 19 Fluid Conditions TAVG 574.2 - 6.5 °F < TAVG - 589.2 + 6.5 OF Pressurizer Pressure 2250 - 30 psia < PRCS < 2250 + 30 psia Reactor Coolant Flow

> 95,700 gpm

" Accumulator Temperature 88 -F - TACC - 120 °F Accumulator Pressure 603 psia - PACC --- 693 psia Accumulator Water Volume 6119 gal -5 VAcc - 6597 gal Accumulator Boron Concentration

Ž 2300 ppm Accident Boundary Conditions Single Failure Assumptions Loss of one ECCS train Safety Injection Flow Minimum Safety Injection Temperature 40 OF 5 Ts, -5 120 OF

" Safety Injection Initiation Delay

_< 17 sec (with offsite power)

Time

_< 27 sec (without offsite power)

Containment Pressure Bounded (minimum) to TXX-07107 Page 9 of 28 Table 2 - Comanche Peak Unit 1 Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement Value Criteria 95/95 PCT (-F) 1492

< 2200 95/95 LMO1 (%)

0.23

< 17 95/95 CWO(2) 0.00

< 1 Notes:

1. Local Maximum Oxidation

2.

Core Wide Oxidation Table 3 - Comanche Peak Unit 1 Best Estimate Large Break Sequence of Events for the Limiting PCT Case Event Time (sec)

Start of Transient 0.0 Safety Injection Signal 5.8 PCT Occurs

-10 Accumulator Injection Begins 12.0 End of Blowdown 25.5 Bottom of Core Recovery 30.0 Accumulator Empty 30.5 Safety Injection Begins 32.8 End of Transient 550.0 to TXX-07107 Page 10 of 28 i 'nA

.t I JU 0

A A

U U

1400t 0

1300+

1200-U-

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I I

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I 1111111 I

I I I I I I.

I I I I I I I.

UU'J 0

0.5 I

1.5 CD

• Abreak/ACL 2

22.5 3

umunoet.x X2007/07/09 MbftI 3MMM Figure 1 - Comanche Peak Unit 1 HOTSPOT PCT versus Effective Break Area Scatter Plot (CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area) to TXX-07107 Page 11 of 28 F-0~

F-0 100 200 300 400 Time After Break (s) 500 1384996879 Figure 2 - Comanche Peak Unit 1 HOTSPOT Clad Temperature Transient at the Limiting Elevation for the Limiting PCT Case to TXX-07107 Page 12 of 28 1600 1400-1200-U-

1000-

<L) 0-E 800-600--

400-200 I1' I

0 100 200 300 400 500 Time After Break (s) 1384996879 Figure 3 - Comanche Peak Unit I HOTSPOT Clad Temperature Transient at the Limiting Elevation for the Limiting LMO Case

Attachment ito TXX-07107 Page 13 of 28 2500 2000-1500-U)

Cl) a)

CO 1000-500-0-

1 I

0 100 200 300 400 500 Time After Break (s) 13M49968879 Figure 4 - Comanche Peak Unit 1 Pressurizer Pressure for the Limiting PCT Case to TXX-07107 Page 14 of 28 70000-60000-50000-(D)

U) 40000-E c*

30000-0 U)

U) S20000-10000-0--

-10000-0 1384996879 100 200 300 400 500 Time After Break (s)

Figure 5 - Comanche Peak Unit 1 Vessel Side Break Flow for the Limiting PCT Case

Attachment ito TXX-07107 Page 15 of 28 INTACT LOOP PUMP 2

INTACT LOOP PUMP 3

INTACT LOOP PUMP 4

BROKEN LOOP PUMP 1

0 0.8 0.6-0 C7, w-0 100 200 300 400 Time After Break (s) 500 1384996879 Figure 6 - Comanche Peak Unit 1 Void Fraction in Pumps for the Limiting PCT Case to TXX-07107 Page 16 of 28 2000 1500-U) 1000-0 I500-0-

-500-0 5

10 15 20 Time After Break (s) 1384996879 Figure 7 - Comanche Peak Unit I Vapor Flow at Top of Core for the Limiting PCT Case to TXX-07107 Page 17 of 28 C)

CI)

E 0

cn Cl) 0 5

10 Time After Break (s) 15 20 13M4996879 Figure 8 - Comanche Peak Unit 1 Total Flow at Bottom of Core for the Limiting PCT Case to TXX-07107 Page 18 of 28 INTACT LOOP 2

ACCUMULATOR MASS FLOW RATE INTACT LOOP 3

ACCUMULATOR MASS FLOW RATE INTACT LOOP 4

ACCUMULATOR MASS FLOW RATE 3000 2500-2000-Cf)

E 1500-1000-Cn )

C/)

ED 0

20.

40 60 80 Time After Break (S) 100 13W496879 Figure 9 - Comanche Peak Unit 1 Accumulator Injection Flow for the Limiting PCT Case to TXX-07107 Page 19 of 28 INTACT LOOP 2

SI MASS FLOW RATE INTACT LOOP 3

SI MASS FLOW RATE INTACT LOOP 4

SI MASS FLOW RATE 1400 1200-1000-E v

800-0 Li L. 6001 0

100 200 300 400 500 Time After Break (s) 13899879 Figure 10 - Comanche Peak Unit 1 Safety Injection Flow for the Limiting PCT Case

Attachment I to TXX-07107 Page 20 of 28 C-)

0 100 200 300 400 Time After Break (s) 500 1384996879 Figure 11 - Comanche Peak Unit 1 Lower Plenum Collapsed Liquid Level for the Limiting PCT Case

Attachment ito TXX-07107 Page 21 of 28 LIQUID LEVEL IN BROKEN LOOP 1 DOWNCOMER LIQUID LEVEL IN INTACT LOOP 2

DOWNCOMER LIQUID LEVEL IN INTACT LOOP 3

DOWNCOMER LIQUID LEVEL IN INTACT LOOP 4

DOWNCOMER 2-C-)

0 100 200 300 400 Time After Break (s) 500 1384998879 Figure 12 - Comanche Peak Unit 1 Downcomer Collapsed Liquid Levels for the Limiting PCT Case

Attachment ito TXX-07107 Page 22 of 28 COLLAPSED LIQUID LEVEL IN LOW POWER CHANNEL 12 COLLAPSED LIQUID LEVEL IN AVERAGE CHANNEL 13 COLLAPSED LIQUID LEVEL IN GUIDE TUBE CHANNEL 14 COLLAPSED LIQUID LEVEL IN HOT ASSEMBLY CHANNEL 15 12 10"

• )

C-)

a.)

0 0

100 200 300 400 Time After Break (s) 500 1384998879 Figure 13 - Comanche Peak Unit I Core Collapsed Liquid Levels for the Limiting PCT Case to TXX-07107 Page 23 of 28 250000--

200000-150000-E U')

U) 100000-50000-0 0

1384996879 100 200 300 400 500 Time After Break (s)

Figure 14 - Comanche Peak Unit I Vessel Fluid Mass for the Limiting PCT Case to TXX-07107 Page 24 of 28 HOT ROD HOT ASSEMBLY GUIDE TUBES OH/SC/OP LOW POWER C-E 0

100 200 300 400 Time After Break (s) 500 1384996879 Figure 15 - Comanche Peak Unit 1 WCOBRAITRAC Peak Clad Temperature for all 5 Rod Groups for the Limiting PCT Case to TXX-07107 Page 25 of 28 0~

C-)

0 100 200 300 400 Time After Break (s) 500 1

384996879 Figure 16 - Comanche Peak Unit 1 Average Core Collapsed Liquid Level per Assembly for the Limiting PCT Case to TXX-07107 Page 26 of 28 0.

0 100 200 300 400 Time After Break (s) 500 1384996879 Figure 17 - Comanche Peak Unit 1 Peak Clad Temperature Elevation for the Hot Rod for the Limiting PCT Case

Attachment Ito TXX-07107 Page 27 of 28 COCO Results WC/T Input Deck Vl) c)30' Cl) 0 100 200 300 400 Time After Break (s) 500 98&395072 Figure 18 - Comanche Peak Unit I Analysis Versus Calculated Containment Backpressure to TXX-07107 Page 28 of 28 0.45 0.40 0.35 I-0 0-0.30 0.25 0.20 4-0.25 0.30 0.35 0.40 PMID 0.45 Figure 19 - Comanche Peak Unit I BELOCA Analysis Axial Power Shape Operating Space Envelope PBOT = integrated power fraction in the bottom third of the core PMID = integrated power fraction in the middle third of the core to TXX-07107 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 2 to TXX-07107 Page 1 of 28 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNIT 2

Background

A best-estimate loss of coolant accident analysis has been completed for Comanche Peak Unit 2. This license amendment request (LAR) for operating license NPF-89 (Comanche Peak Unit 2) requests approval to apply the Westinghouse best-estimate large break loss of coolant accident (LOCA) analysis methodology.

Westinghouse obtained generic NRC approval of its original topical report describing best-estimate large break LOCA methodology in 1996. NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report [1]. This methodology was later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report [2].

Westinghouse recently completed a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 9 5 th percentile. This method is still based on the Code Qualification Document (CQD) methodology [1],[2] and follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics.

The Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-1 6009-P-A [3].

This report summarizes the application of the Westinghouse ASTRUM BELOCA evaluation model to Comanche Peak Unit 2 for the large break LOCA accident analysis. Table 1 lists the major plant parameter assumptions used in the BELOCA analysis for Comanche Peak Unit 2.

Both Luminant Power and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with vendor internal processes for assessing evaluation model changes and errors, are used to identify the need for LOCA analyses impact assessments.

Method of Thermal Analysis for Comanche Peak Unit 2 When the Final Acceptance Criteria (FAC) governing the loss-of-coolant accident (LOCA) for Light Water Reactors was issued in 10 CFR 50.46 [4], both the Nuclear Regulatory Commission (NRC) and the industry recognized that stipulations of Appendix K were highly conservative.

That is, using the then accepted analysis methods, the performance of the Emergency Core Cooling System (ECCS) would be conservatively underestimated, resulting in predicted peak clad temperatures (PCTs) much higher than expected. At that time, however, the degree of to TXX-07107 Page 2 of 28 conservatism in the analysis could not be quantified. As a result, the NRC began a large-scale confirmatory research program with the following objectives:

1.

Identify, through separate effects and integral effects experiments, the degree of conservatism in those models required in the Appendix K rule. In this fashion, those areas in which a purposely prescriptive approach was used in the Appendix K rule could be quantified with additional data so that a less prescriptive future approach might be allowed.

2.

Develop improved thermal-hydraulic computer codes and models so that more accurate and realistic accident analysis calculations could be performed. The purpose of this research was to develop an accurate predictive capability so that the uncertainties in the ECCS performance and the degree of conservatism with respect to the Appendix K limits could be quantified.

Since that time, the NRC and the nuclear industry have sponsored reactor safety research programs directed at meeting the above two objectives. The overall results have quantified the conservatism in the Appendix K rule for LOCA analyses and confirmed that some relaxation of the rule can be made without a loss in safety to the public. It was also found that some plants were being restricted in operating flexibility by overly conservative Appendix K requirements. In recognition of the Appendix K conservatism that was being quantified by the research programs, the NRC adopted an interim approach for evaluation methods. This interim approach is described in SECY-83-472 [5]. The SECY-83-472 approach retained those features of Appendix K that were legal requirements, but permitted applicants to use best-estimate thermal-hydraulic models in their ECCS evaluation model. Thus, SECY-83-472 represented an important step in basing licensing decisions on realistic calculations, as opposed to those calculations prescribed by Appendix K.

In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, "ECCS Evaluation Models", to permit the use of a realistic evaluation model to analyze the performance of the ECCS during a hypothetical LOCA. This decision was based on an improved understanding of LOCA thermal-hydraulic phenomena gained by extensive research programs.

Under the amended rules, best estimate thermal-hydraulic models may be used in place of models with Appendix K features. The rule change also requires, as part of the LOCA analysis, an assessment of the uncertainty of the best estimate calculations. It further requires that this analysis uncertainty be included when comparing the results of the calculations to the prescribed acceptance criteria of 10 CFR 50.46. Further guidance for the use of best estimate codes is provided in Regulatory Guide 1.157 [6].

To demonstrate use of the revised ECCS rule, the NRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology [7].

This method outlined an approach for defining and qualifying a best estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis.

to TXX-07107 Page 3 of 28 A LOCA evaluation methodology for three and four loop pressurized water reactor (PWR) plants based on the revised 10 CFR 50.46 rules was developed by Westinghouse with the support of EPRI and Consolidated Edison and has been approved by the NRC [1]. This methodology was later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report [2].

More recently, Westinghouse developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical TReatment of Uncertainty Method [3]. This method is still based on the CQD methodology and follows the steps in the CSAU methodology.

However, the uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method in which the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing calculations in WCAP-16009-P-A [3]. The ASTRUM methodology remains applicable to three and four loop PWRs, as well as 2-loop Westinghouse plants with UPI. This methodology was also extended to CE design PWRs.

The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95th percentile of the Peak Clad Temperature (PCT), Local Maximum Oxidation (LMO), and Core Wide Oxidation (CWO) with 95% confidence level. These parameters are needed to satisfy the 10 CFR 50.46 criteria with regard to PCT, LMO, and CWO.

Downcomer boiling is considered as appropriate in the ASTRUM methodology. The WCOBRA/TRAC computer code determines if downcomer boiling will occur for a particular transient. If downcomer boiling is determined to occur in a transient, WCOBRA/TRAC includes the effects of downcomer boiling in the transient calculation.

This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A [3] as applicable to the ASTRUM methodology. Section 13-3 of WCAP-1 6009-P-A [3] was found to acceptably disposition each of the identified conditions and limitations related to WCOBRA/TRAC and the CQD uncertainty approach per Section 4.0 of the ASTRUM Final Safety Evaluation Report appended to this WCAP. The Best Estimate LBLOCA analysis and associated model for Comanche Peak Unit 2 is unit-specific.

Description of a Large Break LOCA Transient Before the break occurs, the RCS (Reactor Coolant System) is assumed to be operating normally at full power in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. A large break is assumed to open instantaneously in one of the main RCS cold leg pipes.

Immediately following the cold leg break, a rapid system depressurization occurs along with a core flow reversal due to a high discharge of sub-cooled fluid into the broken cold leg and out of the break. The fuel rods go through departure from nucleate boiling (DNB) and the cladding rapidly heats up, while the core power decreases due to voiding in the core. The hot water in to TXX-07107 Page 4 of 28 the core, upper plenum, and upper head flashes to steam, and subsequently the cooler water in the lower plenum and downcomer begins to flash. Once the system has depressurized to the accumulator pressure, the accumulator begins to inject cold borated water into the intact cold legs. During the blowdown period, a portion of the injected Emergency Core Cooling System (ECCS) water is calculated to be bypassed around the downcomer and out of the break. The bypass period ends as the system pressure continues to decrease and approaches the containment pressure, resulting in reduced break flow and consequently, reduced core flow.

As the refill period begins, the core continues to heat up as the vessel begins to fill with ECCS water.

This phase continues until the lower plenum is filled, the bottom of the core begins to reflood, and entrainment begins.

During the reflood period, the core flow is oscillatory as ECCS water periodically rewets and quenches the hot fuel cladding, which generates steam and causes system re-pressurization. The steam and entrained water must pass through the vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out of the break. This flow path resistance is overcome by the downcomer water elevation head, which provides the gravity driven reflood force. The pumped upper plenum and cold leg injection ECCS water aids in the filling of the vessel and downcomer, which subsequently supplies water to maintain the core and downcomer water levels and complete the reflood period.

ASTRUM Analysis Results for Comanche Peak Unit 2 The results of the Comanche Peak Unit 2 ASTRUM analysis are summarized in Table 2.

Table 3 contains a sequence of events for the limiting PCT transient.

The scatter plot presented in Figure 1 shows the effect of the effective break area on the analysis PCT, The effective break area is calculated by multiplying the discharge coefficient CD with the sample value of the break area, normalized to the cold-leg cross sectional area.

Figure 1 is provided to show the break area is a significant contributor to the variation in PCT.

From the 124 calculations performed as part of the ASTRUM analysis, different cases proved to be the limiting PCT and limiting LMO transient for Comanche Peak Unit 2. Figure 2 shows the predicted clad temperature transient at the PCT limiting elevation for the limiting PCT case.

Figure 3 presents the clad temperature transient predicted at the LMO elevation for the limiting LMO case. Due the low PCT results, CWO remained on the order of 0 percent for all cases.

Figure 4 through 17 illustrate the key major response parameters for the limiting PCT transient.

The reference point for the lower plenum liquid level presented in Figure 11 is the bottom of the vessel (10.1 feet below the bottom of the active fuel). The reference point for the downcomer liquid level presented in Figure 12 is the point at which the outside of the core barrel, if extended downward, intersects with the vessel wall (6.1 feet below the bottom of the active fuel). The reference point for the core collapsed liquid levels presented in Figures 13 and 16 is the bottom of the active fuel.

to TXX-07107 Page 5 of 28 The containment backpressure utilized for the LBLOCA analysis compared to the calculated containment backpressure is provided in Figure 18. The worst single failure for the LBLOCA analysis is the loss of one train of ECCS injection (consistent with the ASTRUM Topical);

however, all containment systems which would reduce containment pressure are modeled for the LBLOCA containment backpressure calculation.

10 CFR 50.46 Requirements It must be demonstrated that there is a high level of probability that the following limits set forth in 10 CFR 50.46 are met:

(b)(1)

The limiting PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT for the limiting case is 16320F for Comanche Peak Unit 2, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., "Peak Clad Temperature less than 2200°F," is demonstrated. The result is shown in Table 2.

(b)(2) The maximum cladding oxidation corresponds to a bounding estimate of the 95th percentile LMO at the 95-percent confidence level. Since the resulting LMO for the limiting case is 0.71 percent for Comanche Peak Unit 2, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Local Maximum Oxidation of the cladding less than 17 percent," is demonstrated. The result is shown in Table 2.

(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. While the limiting LMO is determined based on the single Hot Rod, the CWO value can be conservatively chosen as that calculated for the limiting Hot Assembly Rod (HAR) when there is significant margin to the regulatory limit. The limiting HAR total maximum oxidation is 0 percent for Comanche Peak Unit 2. Thus, a detailed CWO calculation is not needed because the calculations would include many lower power assemblies and the outcome would always be less than the limiting HAR total maximum oxidation. Therefore, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation less than 1 percent," is demonstrated. The result is shown in Table 2.

(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The grid crush calculations currently in place for Comanche Peak Unit 2 remain unchanged with the application of the ASTRUM methodology [3], therefore, acceptance criterion (b)(4) is satisfied.

to TXX-07107 Page 6 of 28 (b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS. Long-term cooling is dependent on the demonstration of continued delivery of cooling water to the core. The actions, automatic or manual, that are currently in place at Comanche Peak Unit 2 to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology [3].

Based on the ASTRUM analysis results (see Table 2), it is concluded that Comanche Peak Unit 2 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

to TXX-07107 Page 7 of 28 References

[1]

Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-1 2945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).

[2]

Dederer, S. I., et. al., 1999, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A, Revision 1 and WCAP-14450, Revision 1 (Non-Proprietary).

[3]

Nissley, M. E., et.al., 2005, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

WCAP-16009-P-A and WCAP-16009-NP-A (Non-Proprietary).

[4]

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 4, 1974.

[5]

Information Report from W.J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods", SECY-83-472, November 17, 1983.

[6]

"Best Estimate Calculations of Emergency Core Cooling System Performance,"

Regulatory Guide 1.157, USNRC, May 1989.

[7]

Boyack, B., et. al., 1989, "Quantifying Reactor Safety Margins: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-of-Coolant-Accident," NUREG/CR-5249.

to TXX-07107 Page 8 of 28 Table I - Major Plant Parameter Assumptions Used in the BELOCA Analysis for Comanche Peak Unit 2 Parameter Value Plant Physical Description SG Tube Plugging

< 10%

Plant Initial Operating Conditions Reactor Power

< 3612 MWt (+/- 0.6% uncertainties)

Peaking Factors FQ < 2.5 FAHN-1.60 Axial Power Distribution See Figure 19 Fluid Conditions TAVG 574.2 - 6.5 F :5 TAVG - 589.2 + 6.5 OF Pressurizer Pressure 2250 - 30 psia < PRCS < 2250 + 30 psia Reactor Coolant Flow

> 95,700 gpm

" Accumulator Temperature 88 F -< TACC - 120 OF Accumulator Pressure 603 psia < PACC < 693 psia Accumulator Water Volume 6119 gal < VACC - 6597 gal Accumulator Boron Concentration

> 2300 ppm Accident Boundary Conditions Single Failure Assumptions Loss of one ECCS train Safety Injection Flow Minimum Safety Injection Temperature 40 °F < Ts, < 120 OF Safety Injection Initiation Delay

< 17 sec (with offsite power)

Time

< 27 sec (without offsite power)

Containment Pressure Bounded (minimum) to TXX-07107 Page 9 of 28 Table 2 - Comanche Peak Unit 2 Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement Value Criteria 95/95 PCT (-F) 1632

< 2200 95/95 LMO( (%)

0.71

< 17 95/95 CWO2 (%)

0.00

< 1 Notes:

1. Local Maximum Oxidation

2.

Core Wide Oxidation Table 3 - Comanche Peak Unit 2 Best Estimate Large Break Sequence of Events for the Limiting PCT Case Event Time (sec)

Event 0.0 Start of Transient 5.3 Safety Injection Signal 10.5 Accumulator Injection Begins 24.0 End of Blowdown 31.0 Bottom of Core Recovery 32.3 Safety Injection Begins 33.1 Accumulator Empty

-85 PCT Occurs 550.0 End of Transient to TXX-07107 Page 10 of 28 i onl

'I I

IUUU M

1600-U 1400-A U-F-

C-)

0~

1200t 4

A AA k

MA HA M

M A

A AAAt AA A

U 1000-I I

I I

I I

III I

I I

I I

I I

I I

I I

I I

I I

I I

OWU 0

0.5 1

1.5 CD

• Abreak/ACL 2

2.5 3

mumpostU X2=0/07/09 20ao~3M 1 3075M~6 Figure 1 - Comanche Peak Unit 2 HOTSPOT PCT versus Effective Break Area Scatter Plot (CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area) to TXX-07107 Page 11 of 28 1,

1 U-a, a,

0~

E a,

0 100 200 300 400 Time After Break (s) 500 1846186506 Figure 2 - Comanche Peak Unit 2 HOTSPOT Clad Temperature Transient at the Limiting Elevation for the Limiting PCT Case to TXX-07107 Page 12 of 28 1600 1400-1200-1000-(D.

ca, E

800-600-400-200

'I'

'I 0

100 200 300 400 500 lTime After Break (s)

Figure 3 - Comanche Peak Unit 2 HOTSPOT Clad Temperature Transient at the Limiting Elevation for the Limiting LMO Case to TXX-07107 Page 13 of 28 2500 2000-1500-U)

U)

Ur) 03 ca_

1000-500-0, 0

100 200 300 400 500 Time After Break (s) 18418M506 Figure 4 - Comanche Peak Unit 2 Pressurizer Pressure for the Limiting PCT Case to TXX-07107 Page 14 of 28 70000--

60000-50000-

_40000--

E

-~30000-0 Cl) 20000-(

n 10000-0-

-10000-0 1846186506 100 200 300 400 500 Time After Break (s)

Figure 5 - Comanche Peak Unit 2 Vessel Side Break Flow for the Limiting PCT Case to TXX-07107 Page 15 of 28 INTACT LOOP PUMP 2

INTACT LOOP PUMP 3

INTACT LOOP PUMP 4

BROKEN LOOP PUMP 1

0.8 0.6-0 C:)

0.2 0

100 20(

Time 18461 86506 0

300 400 500 After Break (s)

Figure 6 - Comanche Peak Unit 2 Void Fraction in Pumps for the Limiting PCT Case to TXX-07107 Page 16 of 28 2000-1500-C-)

CL) 1000-E 0

U 500 0-

-500-1 0

5 10 15 20 Time After Break (s) 1846186506 Figure 7 - Comanche Peak Unit 2 Vapor Flow at Top of Core for the Limiting PCT Case to TXX-07107 Page 17 of 28 40000 30000-20000-C/)

E

-1 10000-0 0-

-10000-

-20000-

/

-30000-0 5

10 15 20 Time After Break (s) 1846186506 Figure 8 - Comanche Peak Unit 2 Total Flow at Bottom of Core for the Limiting PCT Case to TXX-07107 Page 18 of 28 INTACT LOOP 2

ACCUMULATOR MASS FLOW RATE INTACT LOOP 3

ACCUMULATOR MASS FLOW RATE INTACT LOOP 4

ACCUMULATOR MASS FLOW RATE 2500 2000-1500-E 1

1000-0 C,)

0 20 40 60 80 Time After Break (s) 100 1801 88500 Figure 9 - Comanche Peak Unit 2 Accumulator Injection Flow for the Limiting PCT Case to TXX-07107 Page 19 of 28 INTACT LOOP 2

SI MASS FLOW RATE INTACT LOOP 3

SI MASS FLOW RATE INTACT LOOP 4

SI MASS FLOW RATE 1400 1200-1000-E S800-

-)

0 6

001 0

100 200 300 400 500 Time After Break (s) 1840186506 Figure 10 - Comanche Peak Unit 2 Safety Injection Flow for the Limiting PCT Case to TXX-07107 Page 20 of 28 a,

-J a,

0~

C-)

0 100 200 300 400 Time After Break (s) 500 18461865O6 Figure 11 - Comanche Peak Unit 2 Lower Plenum Collapsed Liquid Level for the Limiting PCT Case to TXX-07107 Page 21 of 28 LIQUID LEVEL IN LIQUID LEVEL IN LIQUID LEVEL IN LIQUID LEVEL IN BROKEN LOOP 1 DOWNCOMER INTACT LOOP 2

DOWNCOMER INTACT LOOP 3

DOWNCOMER INTACT LOOP 4

DOWNCOMER a,

a,

-J cj-

-tJ a,cr~

0~

0 C-)

0 100 200 300 400 500 Time After Break (s) 1846186506 Figure 12 - Comanche Peak Unit 2 Downcomer Collapsed Liquid Levels for the Limiting PCT Case to TXX-07107 Page 22 of 28 COLLAPSED LIQUID LEVEL IN LOW POWER CHANNEL 12 COLLAPSED LIQUID LEVEL IN AVERAGE CHANNEL 13 COLLAPSED LIQUID LEVEL IN GUIDE TUBE CHANNEL 14 COLLAPSED LIQUID LEVEL IN HOT ASSEMBLY CHANNEL 15 2-J Cl) 0~

0 100 200 300 400 Time After Break (s) 500 18461885O6 Figure 13 - Comanche Peak Unit 2 Core Collapsed Liquid Levels for the Limiting PCT Case to TXX-07107 Page 23 of 28 250000 200000-150000-E cr*

100000-0 100 200 300 400 500 Time After Break (s) 1840188506 Figure 14 - Comanche Peak Unit 2 Vessel Fluid Mass for the Limiting PCT Case to TXX-07107 Page 24 of 28 HOT ROD HOT ASSEMBLY

---

GUIDE TUBES OH/SC/OP LOW POWER 1600 1400--

1200-1000-Z3 L.

800-L E

600 40-

-1/

600-

/

/

4o-i, 1

£A, 1

200-0 100 200 300 400 500 Time After Break (s) 1846186506 Figure 15 - Comanche Peak Unit 2 WCOBRA/TRAC Peak Clad Temperature for all 5 Rod Groups for the Limiting PCT Case to TXX-07107 Page 25 of 28 0~

0 100 200 300 400 Time After Break (s) 500 1801885O6 Figure 16 - Comanche Peak Unit 2 Average Core Collapsed Liquid Level per Assembly for the Limiting PCT Case to TXX-07107 Page 26 of 28 1.2 6'

cLj 0

100 200 300 400 Time After Break (s) 500 1861 86506 Figure 17 - Comanche Peak Unit 2 Peak Clad Temperature Elevation for the Hot Rod for the Limiting PCT Case to TXX-07107 Page 27 of 28 COCO Results WC/T Input Deck Cl)

Ur)

V/)

a, 0n 0

100 200 300 400 Time After Break (s) 500 1140958353 Figure 18 - Comanche Peak Unit 2 Analysis Versus Calculated Containment Backpressure to TXX-07107 Page 28 of 28 0.45 0.40 0.35 I--

0 0.

0.30 0.25 0.20 ý-

0.25 0.30 0.35 0.40 PMID 0.45 Figure 19 - Comanche Peak Unit 2 BELOCA Analysis Axial Power Shape Operating Space Envelope PBOT = integrated power fraction in the bottom third of the core PMID = integrated power fraction in the middle third of the core to TXX-07107 APPLICATION OF WESTINGHOUSE SMALL BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNITS 1 AND 2 to TXX-07107 Page 1 of 46 APPLICATION OF THE WESTINGHOUSE SMALL BREAK LOCA METHODOLOGY TO COMANCHE PEAK UNITS 1 AND 2 Introduction The small break loss-of-coolant-accident (SBLOCA) analyses for Comanche Peak Units 1 and 2 were completed using the 1985 Westinghouse SBLOCA Evaluation Model with NOTRUMP (NOTRUMP-EM) (References 1, 2 and 3) as a part of the Comanche Peak "Transition Project."

The NOTRUMP-EM SBLOCA analyses were performed at 100.6% of the up-rated core power of 3612 MWt (NSSS power of 3628 MWt) for both units. The Unit 1 NOTRUMP model included the A-76 Replacement Steam Generator (RSG). The purpose of this analysis is to demonstrate conformance with the 10 CFR 50.46 (Reference 4) requirements at the up-rated conditions with NOTRUMP-EM. Important input assumptions, as well as analytical models and analysis methodology for the small break LOCA are contained in subsequent sections. Analysis results are provided in the form of tables and figures. The analysis has shown that no design or regulatory limit related to the small break LOCA would be exceeded at the conditions analyzed.

Input Parameters and Assumptions The SBLOCA methodology using the NOTRUMP-EM was developed in accordance with the requirements of 10 CFR 50 Appendix K. This regulation was designed to produce a conservative prediction of the analysis results and includes various conservative modeling requirements such as the decay heat model (1971 ANS Infinite + 20%), the zirconium-water reaction model (Baker-Just) and the most limiting single failure criterion. For the SBLOCA analysis, loss of offsite power (LOOP) is assumed, which results in the limiting single failure assumption of the loss of one Emergency Diesel Generator (EDG) and a subsequent loss of one train of pumped Emergency Core Cooling System (ECCS). The SBLOCA analysis assumes that reactor trip occurs coincident with the LOOP, which results in the following: (a)

Reactor Coolant Pump (RCP) trip and coastdown and (b) Steam Dump System being inoperable. Additional input assumptions and initial conditions for the SBLOCA analysis are found in Tables 1 - 4 for Units 1 and 2.

Description of Analyses and Calculations Analytical Model The requirements for an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50. For LOCAs due to small breaks, less than 1 square foot in area, the Westinghouse NOTRUMP Small Break LOCA Emergency Core Cooling System (ECCS) Evaluation Model (NOTRUMP-EM, References 1, 2 and 3) is used. NOTRUMP-EM was developed to determine the RCS response to design basis Small Break LOCAs, and to address NRC concerns expressed in NUREG-0611 (Reference 5).

to TXX-07107 Page 2 of 46 NOTRUMP-EM consists of the NOTRUMP and LOCTA-IV computer codes. The NOTRUMP code is employed to calculate the transient depressurization of the Reactor Coolant System (RCS), as well as to describe the mass and energy release of the fluid flow through the break.

Among the features of the NOTRUMP code are: calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, regime-dependent drift flux calculations in multiple-stacked fluid nodes and regime-dependent heat transfer correlations. These features provide NOTRUMP with the capability to accurately calculate the mass and energy distribution throughout the RCS during the course of a small break LOCA.

The RCS model is nodalized into volumes interconnected by flow paths. The broken loop is modeled explicitly, while the intact loops are lumped together into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multi-node capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a calculation of the behavior of the loop seal during a small break LOCA. The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.

Fuel cladding thermal analyses are performed with a version of the LOCTA-IV code (Reference 2) using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions. The LOCTA-IV code models the hot rod and the average hot assembly rod, assuming a conservative power distribution that is skewed to the top of the core. Figure 1 illustrates the code interface for the NOTRUMP-EM.

Analysis Method The small break LOCA analyses considered five different break cases each for Comanche Peak Unit 1 and Unit 2 as indicated by the results in Tables 5, 6 and 7, 8 respectively. A spectrum of cold leg breaks of equivalent diameters 2, 3, 4 and 6 inches and an accumulator line break of equivalent diameter 8.75 inch were considered. For both Units 1 and 2, the 4-inch break was found to be limiting for peak clad temperature (PCT). For both units the 2, 6 and 8.75 inch breaks resulted in minimal or no core uncovery and therefore PCT information was not calculated for these breaks. Note that intermediate (non-integer) breaks were not considered here since the limiting PCTs for the 4 inch break have considerable margin to the 2200°F limit set forth by the acceptance criteria for SBLOCA analysis.

The most limiting single active failure used for a small break LOCA is that of an emergency power train failure which results in the loss of one complete train of ECCS components. In addition, a Loss-of-Offsite Power (LOOP) is postulated to occur coincident with reactor trip.

This means that credit may be taken for at most one centrifugal charging (CCP) pump, one safety injection (SIP) pump and one residual heat removal (RHR) pump. In the analyses for Comanche Peak Units 1 and 2, one CCP, one SIP and one RHR pump are modeled. The small break LOCA analyses performed for both units model the ECCS flow as being delivered to both the intact and broken loops at the RCS backpressure for breaks smaller than the accumulator to TXX-07107 Page 3 of 46 line diameter (2 inch - 6 inch breaks) and at 0 psig containment pressure for the accumulator line diameter (8.75-inch breaks). For the accumulator line break, no ECCS flow is assumed in the faulted loop and the delivered flow to the intact loop is conservatively calculated based on spilling assumption to 0 psig containment back pressure. SI flows are provided in Tables 3 and 4 for each scenario.

Prior to break initiation, the plant is in a full power (100.6%) equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. Other initial plant conditions used in the analysis are given in Table 1. Subsequent to the break opening, a period of reactor coolant system blowdown ensues in which the heat from fission product decay, the hot reactor internals, and the reactor vessel continues to be transferred to the RCS fluid. The heat transfer between the RCS and the secondary system may be in either direction and is a function of the relative temperatures of the primary and secondary conditions. In the case of continuous heat addition to the secondary during a period of quasi-equilibrium, an increase in the secondary system pressure results in steam relief via the steam generator safety valves.

When a small break LOCA occurs, depressurization of the RCS causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low-pressure reactor trip setpoint, conservatively modeled as 1860 psia, is reached. LOOP is postulated to occur coincident with reactor trip. A safety injection signal is generated when the pressurizer low-pressure safety injection setpoint, conservatively modeled as 1715 psia is reached. Safety injection flow is delayed 22 seconds after the occurrence of the low-pressure condition. This delay accounts for signal processing, diesel generator start up and emergency power bus loading consistent with the loss-of-offsite power coincident with reactor trip, as well as the pump acceleration and valve delays.

The following countermeasures limit the consequences of the accident in two ways:

1.

Reactor trip and borated water injection supplement void formation in causing a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay. No credit is taken in the small break LOCA analysis for the boron content of the injection water. In addition, credit is taken in the small break LOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the reactor trip signal, considering the most reactive RCCA is stuck in the full out position.

A rod drop time of 2.4 seconds was used while also considering an additional 2 seconds for the signal processing delay time. Therefore, a total delay time of 4.4 seconds from the time of reactor trip signal to full rod insertion was used in the small break LOCA analysis.

2.

Injection of borated water provides sufficient flooding of the core to prevent excessive cladding temperatures.

During the earlier part of the small break transient (prior to the postulated loss-of-offsite power coincident with reactor trip), the loss of flow through the break is not sufficient to overcome the to TXX-07107 Page 4 of 46 positive core flow maintained by the reactor coolant pumps. During this period, upward flow through the core is maintained. However, following the reactor coolant pump trip (due to a LOOP) and subsequent pump coastdown, the core mixture level decreases until it reaches the top of hot leg elevation (Figures 2B, 3B, 4B, 5B, 6B, 7B, 8B, 9B, 1OB and 11 B). The core heat transfer mechanisms associated with the small break transient include the break itself, the injected ECCS water, and the heat transferred from the RCS to the steam generator secondary side. Main Feedwater (MFW) is conservatively isolated in 7 seconds (consisting of a 2 second signal delay time and a 5 second main feedwater isolation valve stroke time) following the generation of the pressurizer low-pressure SI signal. Additional makeup water is also provided to the secondary using the auxiliary feedwater (AFW) system. An AFW actuation signal is derived from the pressurizer low-pressure SI signal, resulting in the delivery of AFW system flow 60 seconds after the generation of the SI signal. The heat transferred to the secondary side of the steam generator aids in the reduction of the RCS pressure.

The steam generators and the break provide the principal heat removal mechanism until the steam generation in the core is sufficient to establish a flow path through the low elevation loop seal (loop seal clearing) and out of the break. This results in two-phase and ultimately all steam flow through the break which then becomes the principal heat removal mechanism. The rate of core level draining is then slowed as vapor is now allowed to enter the hot legs due to the loop seal clearing. Consistent with the NOTRUMP methodology described in Reference 2, only the faulted loop seal is allowed to clear for the cold leg breaks less than 6 inch equivalent diameter (2 - 4 inches) and both the faulted and the intact loop seals were allowed to clear for breaks

> 6 inch equivalent diameter (6 and 8.75 inches). When the core mixture level drops below the bottom of the hot legs, the rate of uncovery once again establishes itself. The RCS continues to depressurize and the core level continues to decrease.

For the 3 and 4 inch breaks the top of the core uncovers (Figures 2B, 3B, 5B and 9B) leading to an increase in the core exit vapor temperature (Figures 2F and 3F) and the start of clad heat up (Figures 2G, 3G, 5C and 9C). The peak cladding temperature (PCT) occurs near the time when the core is most deeply uncovered and the top of the core is being cooled by steam only.

Minimal or no core uncovery occurs for the 2, 6 and 8.75 inch breaks (Figures 4B, 6B, 7B, 8B, 10B and 11 B) and therefore no clad heat up calculation using the SBLOCTA code is performed for these cases. The safety injection flow rate continues to increase as the RCS pressure decreases. The accumulators begin to inject borated water into the reactor coolant loops (Figures 2E and 3E), when the RCS pressure reaches the accumulator setpoint (including uncertainties) of 603 psia. The safety injection replenishes the core level, which results in a reversal in the clad heat up transient (3 and 4 inch breaks) and a steady increase in the core mixture level. Ultimately, the small break transient analysis is terminated when the~top of the core is recovered or the core mixture level is increasing and ECCS flow provided to the RCS exceeds the break flow rate.

to TXX-07107 Page 5 of 46 Acceptance Criteria The acceptance criteria for the SBLOCA analysis are specified in 10 CFR 50.46, as follows:

1.

The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

2.

The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4.

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5.

After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. (Note that this criterion is not addressed as part of the short-term SBLOCA analysis.)

The acceptance criteria were established to provide a significant margin in ECCS performance following a LOCA.

Results Tables 5 and 7 provide the NOTRUMP transient timing results for Comanche Peak Units 1 and 2 respectively and Tables 6 and 8 provide the SBLOCTA fuel cladding results for Comanche Peak Units 1 and 2 respectively. The peak cladding temperature is 1013 0F for Unit 1 (4 inch break) and 1209°F for Unit 2 (4 inch break), and the maximum local transient oxidation is 0.02 percent for Unit 1 and 0.05 percent for Unit 2. The sum of the pre-transient and transient oxidation remains below 17 percent at all times in life and the average oxidation is negligible for both units.

Limiting Break Case A summary of the transient response for the limiting PCT cases for Unit 1 and Unit 2 are shown in Figures 2A - 2G and Figures 3A - 3G respectively. These figures present the response of the following parameters.

A.

RCS Pressure B.

Core Mixture Level C.

Broken Loop and Intact Loop Pumped Safety Injection Flow Rate to TXX-07107 Page 6 of 46 D.

Break Flow Rate vs. Total ECCS Flow Rate E.

Broken Loop and Intact Loop Accumulator Flow Rate F.

Core Exit Vapor Temperature G.

Hot Rod Clad Temperature and ZrO2 Thickness at PCT Elevation Additional Break Cases Studies documented in Reference 6 have determined that the limiting PCT small break transient occurs for breaks of less than 10-inches in diameter in the cold leg. For Comanche Peak Units 1 and 2, the limiting PCT is captured by the 2, 3, 4, 6 and 8.75 inch break spectrum. Figures 4 (A and B) through 7 (A and B) provide the RCS pressure and Core mixture level plots for the 2, 3, 6 and 8.75 inch breaks for Unit 1. In addition Figure 5C provides the Hot Rod Clad Temperature and ZrO2 Thickness at PCT Elevation for the 3 inch break for Unit 1. Figures 8 (A and B) through 11 (A and B) provide the RCS pressure and Core mixture level plots for the 2, 3, 6 and 8.75 inch breaks for Unit 2. In addition Figure 9C provides the Hot Rod Clad Temperature and ZrO 2 Thickness at PCT Elevation for the 3 inch break for Unit 2. The 2, 6 and 8.75 inch cases resulted in minimal or no core uncovery and therefore PCT information was not calculated.

To summarize, the plots for each of the additional non-limiting break cases include:

A.

RCS Pressure B.

Core Mixture Level C.

Hot Rod Clad Temperature and ZrO 2 Thickness at PCT Elevation (3 inch break only)

Transient Termination The 10 CFR 50.46 criteria continue to be satisfied beyond the end of the calculated transient due to the presence of the following conditions:

1.

The RCS pressure is gradually decreasing.

2.

The net mass inventory is increasing.

3.

The core mixture level is recovered, or recovering due to increasing mass inventory.

4.

As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel cladding temperatures will continue to decline indicating that the temperature excursion is terminated.

Conclusions The small break LOCA analyses considered a break spectrum of 2, 3, 4, and 6 inch equivalent diameter cold leg breaks and a 8.75 inch equivalent diameter accumulator line break for Comanche Peak Units 1 and 2. The analyses presented in this section show that the to TXX-07107 Page 7 of 46 accumulator and safety injection subsystems of the Emergency Core Cooling System, together with the heat removal capability of the steam generator, provide sufficient core heat removal capability to maintain the calculated peak cladding temperatures for small break LOCA below the 2200°F limit of 10 CFR 50.46. Furthermore, the analyses show that the maximum local oxidation (pre-transient plus transient) is less than 17 percent; the core-wide hydrogen generation is less than 1 percent; and the core geometry remains amenable to cooling.

References

1.

Meyer, P. E., "NOTRUMP - A Nodal Transient Small Break and General Network Code,"

WCAP-1 0079-P-A, (proprietary) and WCAP-1 0080-NP-A (non-proprietary),

August 1985.

2.

Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (proprietary) and WCAP-10081-NP-A (non-proprietary), August 1985.

3.

Thompson, C. D. et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-10054-P-A, Addendum 2, Rev. 1 (proprietary), July 1997.

4.

"Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974, as amended in Federal Register, Volume 53, September 1988.

5.

"Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plant," NUREG-0611, January 1980.

6.

Rupprecht, S. D. et al., "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP-1 1145-P-A (proprietary),

October 1986.

to TXX-07107 Page 8 of 46 Table I Input Parameters Used in the Small Break LOCA Analysis (Unit I and Unit 2)

S Unit I Unit 2 A. Core Parameters 100% Licensed Core Power (MWt) 3612 Core Power Calorimetric Uncertainty, %

0.6 Fuel Type 17x17 Vantage +

Total Core Peaking Factor, Fa 2.50 Hot Rod Enthalpy Rise Peaking Factor, FDH 1.60 Hot Assembly Peaking Factor, PHA 1.4245 Axial Offset, %

13 K(z) Limit 2 line segment B. Reactor Coolant System Thermal Design Flow, gpm/loop 95,700 Total Core Bypass Flow, %

5.8 Nominal Vessel Average Temperature Range, OF Max: 589.2 Min: 574.2 Vessel Average Temperature Uncertainty

+/- 60 Pressurizer Pressure (plus uncertainties), psia 2280 Reactor Coolant Pump Type Model 93A 7000HP with Weir Reactor Coolant Pump Weir Height, ft 0.4167 C. Reactor Protection System Reactor Trip Setpoint, psia 1860 Reactor Trip Signal Processing Time (includes Rod 4.4 Drop Time), sec D. Auxiliary Feedwater System Maximum AFW Temperature, OF 1

120 AFW Flow (Minimum) to all 4 Steam Generators, 430 (107.5gpm/SG

• 4)

Initiation Signal Pressurizer Low-Pressure Safety AFW Delivery Delay Time, sec 60 E. Steam Generators Steam Generator Secondary Mass, lbm/SG 101,702 92,000 Steam Generator Tube Plugging, %

10 MFW Isolation Signal Pressurizer Low-Pressure Safety MFW Isolation Delay Time, sec 2

MFW Flow Coastdown Time, sec 5

Feedwater Temperature, OF 450.3 F. Safety Injection SI Water Temperature, OF 120 Safety Injection Signal, psia1 1715 SI Flow Delay Time, sec 22 to TXX-07107 Page 9 of 46 Table 1 Input Parameters Used in the Small Break LOCA Analysis (Unit I and Unit 2)

(cont.)

G. Accumulators Maximum Initial Temperature, OF 120 Initial Water Volume, ft3 850 Minimum Cover Gas Pressure (including 603 uncertainties), psia H. RWST Draindown Input Maximum Containment Spray Flow (2 Trains), gpm 16,000 Minimum Usable RWST Volume, gal 440,300 Table 2 Steam Generator Safety Valve Flows Per Steam Generator (Unit I and Unit 2)

Set Pressure Uncertainty Accumulatio Rated Flow at Full Open Pressure MSSV psig n %

Ibm/hr 1

1185 0

3 893,160 2

1195 0

3 900,607 3

1205 0

3 908,055 4

1215 0

3 915,502 5

1235 0

3 930,397 to TXX-07107 Page 10 of 46 Table 3 Safety Injection Flows for 2-inch to 6-Inch Break Sizes (Spilling to RCS Pressure)

(Unit I and Unit 2)

RCS Pressure Injected Flow Spilled Flow (psia)

(Ibm/s)

(Ibm/s) 14.7 527.37 199.93 34.7 501.77 190.22 114.7 375.32 142.33 134.7 330.65 125.43 154.7 274.99 104.37 174.7 190.59 72.51 194.7 89.88 31.43 214.7 89.88 31.84 314.7 87.24 31.27 414.7 83.89 30.06 514.7 80.43 28.81 614.7 76.81 27.49 714.7 72.89 26.09 814.7 68.79 24.60 914.7 64.45 23.04 1014.7 59.76 21.32 1114.7 54.37 19.36 1214.7 48.08 17.07 1314.7 40.89 14.45 1414.7 29.56 10.31 1514.7 23.02 7.94 1614.7 21.69 7.47 1714.7 20.26 6.99 1814.7 18.79 6.48 1914.7 17.29 5.96 2014.7 15.72 5.42 2114.7 13.70 4.72 2214.7 11.28 3.88 2314.7 7.45 2.57 2414.7 0.00 0.00 to TXX-07107 Page 11 of 46 Table 4 Safety Injection Flows for 8.75-Inch Break Size (Spilling to Containment Pressure)

(Unit I and Unit 2)

RCS Pressure Injected Flow Spilled Flow (psia)

(Ibm/s)

(Ibm/s) 14.7 527.63 190.65 34.7 346.45 344.63 54.7 267.12 397.75 74.7 180.32 451.99 94.7 83.82 509.29 114.7 83.17 507.64 134.7 82.49 507.78 214.7 79.81 508.30 314.7 76.37 508.98 414.7 72.80 509.64 514.7 68.97 510.28 614.7 65.01 510.94 714.7 60.83 511.60 814.7 56.43 512.28 914.7 51.70 512.97 1014.7 46.47 513.65 1114.7 40.65 514.38 1214.7 32.94 517.74 1314.7 21.01 518.82 1414.7 19.10 519.57 1514.7 17.74 519.44 1614.7 16.34 519.73 1714.7 14.72 520.02 1814.7 12.89 520.28 1914.7 10.95 520.57 2014.7 8.54 522.85 2114.7 5.89 523.24 2214.7 2.35 523.64 2314.7 0.00 524.00 2414.7 0.00 524.00 to TXX-07107 Page 12 of 46 Table 5 NOTRUMP Transient Results (Unit 1)

Event (sec) 2-inch(3) 3-inch 4-inch 6-inch(3) 8.75-inch(3)

Break Initiated 0.0 0.0 0.0 0.0 0.0 Reactor Trip Signal 95.94 20.92 12.17 7.56 6.28 Safety Injection Signal 105.75 30.60 21.25 14.56 8.61 Safety Injection Begins(1 )

127.75 52.60 43.25 36.56 30.61 Loop Seal Clearing Occurs(2) 1370

-535

-315 70,135

-22 Core Uncovery N/A.

-955

-715

-440 N/A Accumulator Injection Begins N/A NWA___

-970

-405

-200 Core Recovery N/A

-3000

-1520

-485 N/A RWST Low Level(5) 1608.5 1600.19 1586.95 1574.98 1474.99 Notes:

1. Safety injection begins 22.0 seconds (SI delay time) after the safety injection signal is

2. Loop seal clearing is considered to occur when the broken loop loop seal vapor flow rate is sustained above 1 Ibm/s.

3. There is no core uncovery for the 2-inch and 8.75-inch cases, and only minimal core uncovers for the 6-inch case.

4. Accumulator actually injects at 3485 for TBX, but this is after the core has recovered and has no bearing on the LOCA recovery.

5. The analysis assumesminimum usable RWST volume (440,300 gal) before the low-1 RWST water level signal for switchover to cold leg recirculation is reached.

Table 6 SBLOCTA BOL Results (Unit 1)

Result 2-inch(2) 3-inch 4-inch 6-inch(2) 8.75-inch(2)

PCT, OF 949.6 1012.8 PCT Time, sec 2026.1 1058.3 PCT Elevation, ft 10.75 11.00 HR Burst Time 1, sec N/A N/A HR Burst Elevation(1 ), ft N/A N/A N/A N/A N/A Max. Local Transient ZrO2, %

0.02 0.01 Max. Local ZrO2 Elevation, ft 10.75 11.00 Average ZrO 2, %

0.00 0.00 Notes:

1, None of the hot rods nor the hot assembly rods burst during the SBLOCTA calculations.

2, The core either does not uncover or only uncovers for a very short time and therefore does not warrant SBLOCTA calculations for these break sizes.

to TXX-07107 Page 13 of 46 Table 7 NOTRUMP Transient Results (Unit 2)

Event (sec) 2-inch(_ )

3-inch 4-inch 6-inch (3) 8.75-inch(3)

Break Initiated 0.0 0.0 0.0 0.0 0.0 Reactor Trip Signal 101.81 35.60 12.52 7.79 6.19 Safety Injection Signal 111.82.

44.53 21.99 14.81 8.58 Safety Injection Begins(1 )

133.82 66.53 43.99 36.81 30.58 Loop Seal Clearing Occurs(2) 1140 510 255 135

-17 Core Uncovery N/A

-670

-575

-420 N/A Accumulator Injection Begins N/A N/A___

-850

-380

-170 Core Recovery N/A

-2800

-1630

-470 N/A RWST Low Level(5) 1609.1 1599.10 1585.37 1574.83 1474.55 Notes:

1. Safety injection begins 22.0 seconds (SI delay time) after the safety injection signal is

2. Loop seal clearing is considered to occur when the broken loop loop seal vapor flow rate is sustained above 1 Ibm/s.

3. There is no core uncovery for the 2-inch and 8.75-inch cases, and only minimal core uncovers for the 6-inch case.

4. Accumulator actually injects at -3330 sec for TCX, but this is after the core has recovered and has no bearing on the LOCA recovery.

Table 8 SBLOCTA BOL Results (Unit 2)

Result 2-inch(2) 3-inch 4-inch 6-inch(2) 8.75-inch(2)

PCT, 'F 1068.5 1209.4 PCT Time, sec 1787.9 919.7 PCT Elevation, ft 11.0 11.0 HR Burst Time(1), sec N/A N/A HR Burst Elevation(1 ), ft N/A N/A N/A N/A N/A Max. Local Transient ZrO2, %

0.04 0.06 Max. Local ZrO 2 Elevation, ft 11.0 11.0 Average ZrO2, %

0.01 0.01 Notes:

1. None of the hot rods nor the hot assembly rods burst during the SBLOCTA calculations.

2.

The core either does not uncover or only uncovers for a very short time and therefore does not warrant SBLOCTA calculations for these break sizes.

to TXX-07107 Page 14 of 46 CORE PRESSURE, CORE FLOW, MIXTURE LEVEL, AND FUEL ROD POWER N

HISTORY O

O<TIME<CORE COVERED L

T R

C U

T M

A P

Figure 1 Code Interface Description for NOTRUMP-EM to TXX-07107 Page 15 of 46 2500 2000-1500-

.2 Q -1000................

~-100........

500-0 I

0 1000 2000 3000 4000 Time (s)

Figure 2A RCS Pressure 4-Inch Break (Unit 1) to TXX-07107 Page 16 of 46 Core Mixture Level

.Top of Core (22.0778 ft) 32 30 28-26-24-0 1000 2000 3000 Time (s) 4W00 Figure 2B Core Mixture Level 4-Inch Break (Unit 1) to TXX-07107 Page 17 of 46 Broken Loop SI Flow Intact Loop SI Flow 100 S.~

/C I

I cn t

E.

60..

I o

2-.....................................................

I I

I I

I I

I I

I I

I II I

I I

0 1000 2000 M00 4MC lime (s)

Figure 2C Broken Loop and Intact Loop Pumped Safety Injection Flow Rate 4-Inch Break (Unit 1)

)0 to TXX-07107 Page 18 of 46 Total Break Flow Total ECCS Flow 2500" 2500.................

0 A.)

0 1000 2000 3000 Time (s) 4000 Figure 2D Break Flow Rate vs. Total ECCS Flow Rate 4-Inch Break (Unit 1) to TXX-07107 Page 19 of 46 Broken Loop Accumulator Flow Intact Loop Accumulator Flow 350-300-..................

250.....................

CO, E

-I II 21 0................

...... 03 0

t i.....

C/)

C,,

M[

0 1000 2000 3000 Time (s) 4000 Figure 2E Broken Loop and Intact Loop Accumulator Flow Rate 4-Inch Break (Unit 1) to TXX-07107 Page 20 of 46 900 800.

700.............................................

500-5-00..............

400 I

I I

I I

I I

I I

I I

I I

I I

I 0

1000 2000 3000 4000 Time (s)

Figure 2F Core Exit Vapor Temperature 4-Inch Break (Unit 1) to TXX-07107 Page 21 of 46 Temperature (F)

Hot Rod Clad Average Temperaure at PCT Elevation (11.0 ft)

Oxide Thickness

()

Maximum ZRO2 Thickness (11.0 ft)

(-03 C.-a--

E a) 1-1 C,)

500 1000 1500 2000 2500 Time (S 3000 3500 4000 Figure 2G Hot Rod Clad Temperature and ZRO2 Thickness at PCT Elevation 4-Inch Break (Unit 1) to TXX-07107 Page 22 of 46 2500 2000 "1500-500 Uf)

CO 5-000 0

I I

I I

I I

I I

I I

I I

01000 2000 3000 4000 Time (s)

Figure 3A RCS Pressure 4-Inch Break (Unit 2) to TXX-07107 Page 23 of 46 Core Mixture Level Top of Core (22.0778 ft)

-J 0

1000 2000 3000 Time (s) 4W0 Figure 3B Core Mixture Level 4-Inch Break (Unit 2) to TXX-07107 Page 24 of 46 Broken Loop SI Flow Intact Loop SI Flow 100-80 I.

I o

I I

Ct, E

a>

I 6

0 n 0-I I

I I

I I

I I

I I

I I

I I

0 1000 2000 3000 40(

Time (s)

Figure 3C Broken Loop and Intact Loop Pumped Safety Injection Flow Rate 4-Inch Break (Unit 2)

• 0 to TXX-07107 Page 25 of 46 Total Break Flow Total ECCS Flow 250 S2000...................

E S1500................

L 0

1000 2000 3000 Time (s) 4000 Figure 3D Break Flow Rate vs. Total ECCS Flow Rate 4-Inch Break (Unit 2) to TXX-07107 Page 26 of 46 Broken Loop Accumulator Flow Intact Loop Accumulator Flow 500 400............................

r 7 E

..* 300..

0 O 200.

M2.

!i 0

1000 2000 3000 4000 Time (s)

Figure 3E Broken Loop and Intact Loop Accumulator Flow Rate 4-Inch Break (Unit 2) to TXX-07107 Page 27 of 46 900 800...........

LL.

00..

-- 60 0..

500......

...i.....

400 0

1000 2000 3000 4000 Time (s)

Figure 3F Core Exit Vapor Temperature 4-Inch Break (Unit 2) to TXX-07107 Page 28 of 46 Temperature (F)

Hot Rod Clad Average Temperaure at PCT Elevation (11.0 ft)

Oxide Thickness

(,X)

Maximum ZRO2 Thickness (11.0 ft)

E FT C,,

CO C.=

C) 0 1000 2000 3000 400 Time (s)

Figure 3G Hot Rod Clad Temperature and ZrO2 Thickness at PCT Elevation 4-Inch Break (Unit 2) to TXX-07107 Page 29 of 46 2400 2200-.

2000 co 01 00.

qC) 1600-0'3 1400.

120.

1000 0

2000 4000 6000 8000 Time (s)

Figure 4A RCS Pressure 2-Inch Break (Unit 1) to TXX-07107 Page 30 of 46 Core Mixture Level Top of Core (22.0778 ft) 32 30"

~28.............

-J26

~26t.............................

0 2000 4000W 000 Time (s) 8000 Figure 4B Core Mixture Level 2-Inch Break (Unit 1) to TXX-07107 Page 31 of 46 2500-2000 S1500-500 I

I I

4000 0

1000 2000 3000 4000 Time (s)

Figure 5A RCS Pressure 3-Inch Break (Unit 1) to TXX-07107 Page 32 of 46 Core Mixture Level Top of Core (22.0778 ft) 32 30-28-26-24-0 1000 2000 3000 4000 Time (s)

Figure 5B Core Mixture Level 3-Inch Break (Unit 1) to TXX-07107 Page 33 of 46 Temperature (F)

Hot Rod Clad Oxide Thickness (Z)

Maximum ZR 2 1000 900.............

900-U-"*

7oo.

Z700 E

Average Temperaure at PCT Elevation (10.75 ft)

Thickness (10.75 ft)

C:)

20005~

Figure 5C Hot Rod Clad Temperature and ZrO2 Thickness at PCT Elevation 3-Inch Break (Unit 1) to TXX-07107 Page 34 of 46 2500O 2000 1500-

.2.

5-000 0

I I

I I

I I

I I

I3I I I I I I

I3 0 0 0I I

U I

I I

0 500 1000 1500 2000 2500 3000 Time (s)

Figure 6A RCS Pressure 6-Inch Break (Unit 1) to TXX-07107 Page 35 of 46 Core Mixture Level Top of Core (22.0778 ft) 32 30 28 26-24 -..

X..

0 500 1000 1500 2000 2500 Time (s) 3000 Figure 6B Core Mixture Level 6-Inch Break (Unit 1) to TXX-07107 Page 36 of 46 2500-2000 1500 C,,

500" 0

I I

I I

10 I

I I I I I I I I2 II 0I 0

500 1000 1500.

2000 2500 3000 Time (s)

Figure 7A RCS Pressure 8.75-Inch Break (Unit 1) to TXX-07107 Page 37 of 46 Core Mixture Level Top of Core (22.0778 ft) 32 30-

• -2 26 0

500 1000 1500 2000 2500 3000 Time (s)

Figure 7B Core Mixture Level 8.75-Inch Break (Unit 1) to TXX-07107 Page 38 of 46

')Arln v

'-rvU 2200-1 2000-En 1800-U.)

C 16 P

1600-1400-1200-I UUU 0

2000 4000 Time (s) 6000 8000 Figure 8A RCS Pressure 2-Inch Break (Unit 2) to TXX-07107 Page 39 of 46 Core Mixture Level Top of Core (22.0778 ft)

X, 0

2000 4W0 6MW Time (s) 8000 Figure 8B Core Mixture Level 2-Inch Break (Unit 2) to TXX-07107 Page 40 of 46 2500 2000 C 1500.

Co CO 1000..

500120 0

1000 2000 3000 4000 Time (s)

Figure 9A RCS Pressure 3-Inch Break (Unit 2) to TXX-07107 Page 41 of 46 Core Mixture Level Top of Core (22.0778 ft) 32-30-28-"

6-26-202 20 "-

18 0

1000 2000 3000 4z00 Time (s)

Figure 9B Core Mixture Level 3-Inch Break (Unit 2) to TXX-07107 Page 42 of 46 Temperature (F)

Hot Rod Clad Oxide Thickness (M)

Maximum ZR 2 1100 1000-900.

800-E)

E 700..............

Average Temperoure at PCT Elevation (11.0 ft)

Thickness (11.0 ft)

CO)

C-0 FE a)

2 X

0 500 1000 1500 2000 2500 3000 3500 4000 Time (s)

Figure 9C Hot Rod Clad Temperature and ZrO2 Thickness at PCT Elevation 3-Inch Break (Unit 2) to TXX-07107 Page 43 of 46 2500.

2000

.200

-~ 500-1000 500 0

I I

I I

I I

I I

I I

I I3000 u

I I

I 0

500 1000 1500 2000 2500 3000 Time (s)

Figure 10A RCS Pressure 6-Inch Break (Unit 2) to TXX-07107 Page 44 of 46 Core Mixture Level Top of Core (22.0778 It) 32 30........

28-

~24....

X.

9-4.--.'-

-p-P

,-.q 1,.."

0 500 1000 1500 2000 2500 Time (s) 3000 Figure lOB Core Mixture Level 6-Inch Break (Unit 2) to TXX-07107 Page 45 of 46 2500-2000....................................................

~1500..

1000.

500...........

500 0 t t

t ltit t

I I

1 II 0III 2I 5

0III I

I II 0

500 1000 1500 2000 2500 3000 Time (s)

Figure 1 IA RCS Pressure 8.75-Inch Break (Unit 2) to TXX-07107 Page 46 of 46 Core Mixture Level Top of Core (22.0778 ft)

,zq

• JL 30-1......

A 28" 0)

-J 0)

L..,

w UI 261 U,

~..........

2-W% I i

i i

i 1U 0

500 1000 1500 Time (s) 2 2O000 I

2500 3000 Figure IIB Core Mixture Level 8.75-Inch Break (Unit 2)