ET 10-0025, Application to Revise Technical Specification 5.6.5, Core Operating Limits Report (Colr), for Large Break Loss-of-Coolant Accident Analysis Methodology

Application to Revise Technical Specification 5.6.5, Core Operating Limits Report (Colr), for Large Break Loss-of-Coolant Accident Analysis Methodology

ML103200209
Person / Time
Site:	Wolf Creek
Issue date:	11/04/2010
From:	Garrett T Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 10-0025
Download: ML103200209 (43)

Text

WOLF CREEK

'NUCLEAR OPERATING CORPORATION Terry J. Garrett Vice President Engineering November 4, 2010 ET 10-0025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

1) WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January, 2005

2) Letter from H. N. Berkow (USNRC) to J. A. Gresham (Westinghouse Electric Company), "Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" (TAC NO.

MB9483)," November 5, 2004

Subject:

Docket No. 50-482: Application To Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," for Large Break Loss-of-Coolant Accident Analysis Methodology Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The proposed amendment revises Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to replace the existing large break loss-of-coolant accident (LOCA) analysis methodology. Specifically, the proposed change adds a reference to WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," (Reference 1) to TS 5.6.5b. The Nuclear Regulatory Commission (NRC) approved WCAP-16009-P in Reference 2.

Attachment I through V provide the Evaluation, Markup of TSs, Retyped TS pages, proposed TS Bases changes, and proposed COLR changes, respectively, in support of this amendment request. Attachment IV and V, is provided for information only. Final TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specification (TS) Bases Control Program," at the time the amendment is implemented.

" 0 0 !Jhh

),I-1 P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET V,,\ ýJt

ET 10-0025 Page 2 of 3 It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The Plant Safety Review Committee reviewed this amendment application. In accordance with 10 CFR 50.91, a copy of this amendment application, with attachments, is being provided to the designated Kansas State official.

WCNOC requests approval of the proposed amendment by September 29, 2011. It is anticipated that the license amendment, as approved, will be effective upon issuance and will be implemented within 90 days from the date of issuance. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr. Richard D. Flannigan at (620) 364-4117.

Sini Terry J. Garrett TJG/rlt Attachments: Evaluation IV Proposed Technical Specification Changes (Mark-up)

II Revised Technical Specification Pages IV Proposed TS Bases Changes (for information only)

V Proposed COLR Changes (for information only) cc: E. E. Collins (NRC), w/a T. A. Conley (KDHE), w/a G. B. Miller (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a

ET 10-0025 Page 3 of 3 STATE OF KANSAS )

SS COUNTY OF COFFEY )

Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By __ _ _

Terry ýJ/-carre'tt Vice President Engineering SUBSCRIBED and sworn to before me this dayofI/Vemb.r ,2010.

b

[A _GAYLE SHEPHEARD Notary Public - State of Kansas MyADt*t Exoires  ?/1qJ2 Zv ) Ept Date ')/*//'_/

It ~~~Expiration Dt eAI

Attachment I to ET 10-0025 Page 1 of 30 EVALUATION 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusion

6.0 REFERENCES

Attachment I to ET 10-0025 Page 2 of 30 EVALUATION 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The proposed amendment revises Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to replace the existing large break loss-of-coolant accident (LOCA) analysis methodology. Specifically, the proposed change adds a reference to WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," (Reference 1) to TS 5.6.5b. The Nuclear Regulatory Commission (NRC) approved WCAP-16009-P in Reference 2.

2.0 DETAILED DESCRIPTION TS Section 5.6.5a. requires core operating limits to be established and documented in the COLR prior to each reload cycle, or prior to any remaining portion of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, as listed in TS Section 5.6.5b. The current methodology used for development of core operating limits related to the large break LOCA (LBLOCA) is listed in TS Section 5.6.5b.7, which states:

7. WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code."

The proposed change replaces the reference to WCAP-1 0266-P-A with a reference to WCAP-16009-P-A, as follows:

7. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."

Listing only the NRC-approved methodology by topical report number and title is consistent with Amendment No. 144 (Reference 8). In Amendment No. 144, the staff concluded that the proposed change to only list the NRC-approved methodology by topical report number and title is acceptable. Attachment IV provides proposed changes to the COLR that includes listing the NRC-approved methodology including the specific revision of the topical report.

3.0 TECHNICAL EVALUATION

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 (Reference 2).

Westinghouse recently underwent a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is still based on the Code Qualification Document (CQD) methodology (Reference 3) 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 where

Attachment I to ET 10-0025 Page 3 of 30 the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A (Reference 1).

A best estimate large break LOCA analysis was completed for WCGS. The application of the Westinghouse ASTRUM best estimate LOCA evaluation model for the large break LOCA analyses is summarized below. Table 1 lists the major plant parameter assumptions used in the BELOCA analysis for WCGS. Both WCNOC and its analysis vendor (Westinghouse) have interface processes that 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.

Table 1 Major Plant Parameter Assumptions Used in the BELOCA Analysis for WCGS Parameter - Value Plant Physical Description

• SG Tube Plugging <10%

Plant Initial Operating Conditions

• Reactor Power <3565 MWt (+ 2% uncertainty)

• Peaking Factors FH -<2.5

_______________________________FAH *ý 1.65

• Axial Power Distribution See Figure 18 Fluid Conditions

• TAVG 570.7 - 4.0 °F < TAVG-- 588.4 + 4.0 OF

" Pressurizer Pressure 2250 - 50 psia < PRCS -<2250 + 50 psia

" Reactor Coolant Flow > 90,324 gpm/loop

" Accumulator Temperature 50 OF < TAcc < 120 OF

• Accumulator Pressure 582.8 psia _<PACC -<696.6 psia

• Accumulator Water Volume 818 ft3 < VAcc -<881 ft3

" Accumulator Boron > 2300 ppm Concentration Accident Boundary Conditions

" Single Failure Assumptions Loss of one ECCS train

• Safety Injection Flow Minimum

• Safety Injection Temperature 37 0F _<TSI < 120 OF

" Safety Injection Initiation Delay 27 sec. (with offsite power)

Time 39 sec. (without offsite power)

• Containment Pressure Bounded (minimum); see Figure 17

Attachment I to ET 10-0025 Page 4 of 30 3.1 Methodology Background When the final acceptance criteria (FAC) governing the LOCA for light water reactors was issued in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water reactors (Reference 4), both the Nuclear Regulatory Commission (NRC) and the industry recognized that stipulations of Appendix K, "ECCS Evaluation Models," 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 cladding temperatures (PCT's) 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:

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 (Reference 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 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 (Reference 6).

Attachment I to ET 10-0025 Page 5 of 30 To demonstrate use of the revised ECCS rule, the NRC and its consultants developed the CSAU evaluation methodology (Reference 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 (Reference 3) 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 Electric Power Research Institute (EPRI) and Consolidated Edison and has been approved by the NRC.

More recently, Westinghouse developed an alternative uncertainty methodology (Reference 1) called ASTRUM. This method is still based on the CQD methodology and follows the steps in the CSAU methodology. However, the uncertainty analysis (i.e., 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 where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology was approved by the NRC, as documented in the safety evaluation appended to Reference 1.

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

This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A as applicable to the ASTRUM methodology, and the conditions and limitations discussed in Section 4.0 of Reference 2. Section 13-3 of WCAP-16009-P-A 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 appended to this WCAP. A best estimate LBLOCA analysis, and associated model, was completed for WCGS.

3.2 Description of a LBLOCA Transient Before the break occurs, the Reactor Coolant System (RCS) 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 legs. During the blowdown period, a portion of the injected 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.

Attachment I to ET 10-0025 Page 6 of 30 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 downcomerwater elevation head, which provides the gravity driven reflood force. The pumped 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.

3.3 Realistic LBLOCA Analyses Results 3.3.1 ASTRUM Analyses Results The results of the WCGS ASTRUM analysis are summarized in Table 2. Table 3 contains a sequence of events for the limiting PCT transient.

Table 2 WCGS Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement Value Criteria 95/95 PCT (°F) 1576 < 2200 95/95 LMO(1 (%) 0.97 < 17 95/95 CWO(2 ) (%) 0.0 <1 Note: (1)LMO - Local Maximum Oxidation (2) CWO - Core Wide Oxidation

Attachment I to ET 10-0025 Page 7 of 30 Table 3 WCGS Best Estimate Large Break Sequence of Events for the Limiting PCT Case in Seconds Event Time (sec)

Start of Transient 0.0 Safety Injection Signal 5.7 Accumulator Injection Begins 10.5 End of Blowdown 25 Bottom of Core Recovery 32.5 Accumulator Begins to Empty 35.8 Safety Injection Begins 44.7 PCT Occurs 60.6 Quench Time 250 End of Transient 350 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 sampled value of the break area, normalized to the cold-leg cross sectional area.

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

Attachment I to ET 10-0025 Page 8 of 30 PCT vs. (CD, A) (All 124 Cases)

• PCT DEG 0 0 PCT DEGCL ddeg F]

A A PCT - SPL 0 0 0 PCT SPL IT [deg F]

F--

o-m2 0 0.5 1 1.5 2 2.5 CD

• Abre1k/ACL sumpost-x X2009/06/23 11016:06.27 20a3575543 Figure 1 - HOTSPOT PCT vs. Effective Break Area Scatter Plot (non-IFBA)

CD = Discharge Coefficient Abreak = Break Area ACL = Cold Leg Area

Attachment I to ET 10-0025 Page 9 of 30 From the 124 calculations performed as part of the ASTRUM analysis, the same case proved to be the limiting PCT and LMO transient (CWO was predicted to be 0.0% for all cases). Figure 2 shows the predicted HOTSPOT cladding temperature transient at the PCT location for the limiting PCT and LMO case. The HOTSPOT PCT plot includes local uncertainties applied to the Hot Rod. Figure 3 provides the PCT elevation on the Hot Rod for the limiting case. Figure 4 provides the WCOBRA/TRAC PCT for all the fuel rods in the core channels.

HOTSPOT PCT I

E OD 0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 2 - HOTSPOT Cladding Temperature Transient for the Limiting Case

Attachment I to ET 10-0025 Page 10 of 30 PCT-Loca t i on 11-10.5 10 I 95 0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 3 - PCT Elevation for the Hot Rod for the Limiting PCT Case

Attachment I to ET 10-0025 Page 11 of 30 Hot Rod Hot Assembly Rod

--- - -- CGuide Tubes Average Rod OH-SC-OP Average Rod L.Low Power Rod U-Q)

E 800 0) 0 50 100 150 200 250 500 350 Time After Break (s) 1436462808 Figure 4 - WCIT PCT for all 5 Rod Groups for the Limiting PCT Case

Attachment I to ET 10-0025 Page 12 of 30 Figures 5 through 16 illustrate the key major response parameters for the limiting PCT and LMO transient. The containment backpressure utilized for the LBLOCA analysis compared to the calculated containment backpressure is provided in Figure 17. 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 that would reduce containment pressure are modeled for the LBLOCA containment backpressure calculation.

VESSEL SIDE BREAK FLOW Cn E

0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 5 - Vessel Side Break Flow for the Limiting PCT Case

Attachment I to ET 10-0025 Page 13 of 30 PUMP SIDE BREAK FLOW E

CD 0

C/)

0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 6 - Pump Side Break Flow for the Limiting PCT Case

Attachment I to ET 10-0025 Page 14 of 30 Intoct Loop Broken Loop 0-o I,

0 0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 7 - Void Fraction in Reactor Coolant Pumps for the Limiting PCT Case

Attachment I to ET 10-0025 Page 15 of 30 VAPOR FLOW RATE IN CORE HOT ASSEMBLY CHANNEL 15 cl-I E

0) 0*

0 0

0~

ci 0 5 10 15 20 25 30 Time After Break (s) 1436462808 Figure 8 - Vapor Flow in the Hot Assembly Channel for the Limiting PCT Case

Attachment Ito ET 10-0025 Page 16 of 30 PRESSURIZER PRESSURE 2500 2000-1500-CD G)

U) 0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 9 - Pressurizer Pressure for the Limiting PCT Case

Attachment I to ET 10-0025 Page 17 of 30 LOWER PLENUM COLLAPSED LIQUID LEVEL 12 106 ... . . . .. . . . . . .. . .

8--

-C:)

_CJ Cl) 0 0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 10 - Lower Plenum Collapsed Liquid Level for the Limiting PCT Case

Attachment I to ET 10-0025 Page 18 of 30 INTACT LOOP 2 ACCUMULATOR MASS FLOW RATE 2500 2000 1500-Cn E

(D c* 1000-03

.n 0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 11 - Accumulator Injection Flow for the Limiting PCT Case

Attachment I to ET 10-0025 Page 19 of 30 INTACT LOOP 2 CHARGING SI MASS FLOW RATE 1

1 12 U)

- 10-E U)

U) rCD 0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 12 - Charging Safety Injection Flow for the Limiting PCT Case

Attachment I to ET 10-0025 Page 20 of 30 INTACT LOOP 2 SIP+RHR SI MASS FLOW RATE 0)

Cn C/)

0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 13 - SIP + RHR Safety Injection Flow for the Limiting PCT Case

Attachment I to ET 10-0025 Page 21 of 30 COLLAPSED LIQUID LEVEL IN INTACT LOOP 2 DOWNCOMER 35-o-j 0Z)-

.-q

-0C 0 50 100 150 200 250 Time After Break (s) 1436462808 Figure 14 - Downcomer Collapsed Liquid Level for the Limiting PCT Case

Attachment I to ET 10-0025 Page 22 of 30 COLLAPSED LIQUID LEVEL IN AVERAGE CHANNEL 13 1

0 0

-J 0~

0 C/I 0

C-)

0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 15 - Collapsed Liquid Level in Core Average Channel for the Limiting PCT Case

Attachment I to ET 10-0025 Page 23 of 30 VESSEL LIQUID MASS F-i C/I 0 50 100 150 200 250 300 350 Time After Break (s) 1436462808 Figure 16- Vessel Liquid Mass for the Limiting PCT Case

Attachment I to ET 10-0025 Page 24 of 30 Cal culated Containment Backpressu re (COCO) alysi A nra s Conta i nment Backpressure (WC /T)

U.)

E 20 O:3 Q) n 0 50 100 150 200 250 300 350 Time (s) 690332938 Figure 17 - Analysis vs. Calculated Containment Backpressure

Attachment I to ET 10-0025 Page 25 of 30 PBOT/PMID Box For Wolf Creek BELOCA Project 0.50 0.31, 0.45 0.45 00.30,0.43.

0 0.40 -

0.35

.C

. . .. *. . . . .. .. . .... . .. . . . . . . . i . . . . . . . . . . . . . . . . . . . . . . . -.. . . . . . ..-

0.30 i ............. *...f o0 445, 0 26 0.25 0.43, 0.23 0.20 A1C 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 Power in Middle Third of Core (PMID)

Figure 18 - WCGS BELOCA Analysis Axial Power Shape Operating Space Envelope

Attachment I to ET 10-0025 Page 26 of 30 3.3.2 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:

10 CFR 50.46(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 1576 IF for WCGS, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., "Peak Cladding Temperature less than 2200 IF," is demonstrated. The result is shown in Table 2.

10 CFR 50.46(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 0.97 percent for WCGS, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e.,

"Local Maximum Oxidation of the cladding less than 17 percent of the total cladding thickness before oxidation," is demonstrated. The result is shown in Table 2.

10 CFR 50.46(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 WCGS. 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 of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume," is demonstrated. The result is shown in Table 2.

10 CFR 50.46(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 WCGS remain unchanged with the application of the ASTRUM methodology; therefore, acceptance criterion (b)(4) is satisfied.

10 CFR 50.46(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 WCGS to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology.

Attachment I to ET 10-0025 Page 27 of 30 Based on the ASTRUM analysis results (see Table 2), it is concluded that WCGS continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Section 50.46 of Title 10 of the Code of Federal Regulations (10 CFR 50.46), "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," requires, in part, that ECCS cooling performance be calculated in accordance with an acceptable evaluation model and be calculated for a number of postulated LOCAs of different sizes, locations, and other properties. It also requires that "... uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated.

This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded."

Section 50.46(b) of 10 CFR 50 also states detailed acceptance criteria for LOCA evaluations.

These are 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 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.

4.2 Significant Hazards Consideration Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) is requesting an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The proposed amendment revises Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to replace the existing large break loss-of-coolant accident (LOCA) analysis methodology. Specifically, the proposed change adds a reference to WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," to TS 5.6.5b. The Nuclear Regulatory Commission (NRC) approved WCAP-16009-P in a safety evaluation dated November 5, 2004.

WCNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment:

Attachment I to ET 10-0025 Page 28 of 30

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises TS Section 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed change adds WCAP-16009-P-A to TS 5.6.5b. as a method used for establishing core operating limits.

Accident analyses are not accident initiators; therefore, the proposed change does not involve a significant increase in the probability of an accident. The analyses using ASTRUM demonstrated that the acceptance criteria in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for lightwater nuclear power reactors," were met. Large break LOCA analyses performed consistent with the methodology in NRC approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions, demonstrate that 10 CFR 50.46 acceptance criteria are met; thus, this change does not involve a significant increase in the consequences of an accident. No physical changes to the plant are associated with the proposed change.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed change revises TS Section 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed change adds WCAP-16009-P-A to TS 5.6.5b. as a method used for establishing core operating limits. There are no physical changes being made to the plant as a result of using the Westinghouse ASTRUM analysis methodology in WCAP-16009-P-A for performance of the large break LOCA analyses. Large break LOCA analyses performed consistent with the methodology in NRC approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions, demonstrate that 10 CFR 50.46 acceptance criteria are met. No new modes of plant operation are being introduced. The configuration, operation, and accident response of the structures or components are unchanged by use of the new analysis methodology. Analyses of transient events have confirmed that no transient event results in a new sequence of events that could lead to a new accident scenario. The parameters assumed in the analyses are within the design limits of existing plant equipment.

In addition, employing the Westinghouse ASTRUM large break LOCA analysis methodology does not create any new failure modes that could lead to a different kind of accident. The design of systems remains unchanged and no new equipment or systems have been installed which could potentially introduce new failure modes or accident sequences. No changes have been made to instrumentation actuation setpoints. Adding the reference to WCAP-16009-P-A in TS Section 5.6.5b. is an administrative change that does not create the possibility of a new or different kind of accident.

Attachment I to ET 10-0025 Page 29 of 30 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises TS Section 5.6.5 to incorporate a new large break LOCA analysis methodology. Specifically, the proposed change adds WCAP-1 6009-P-A to TS 5.6.5b. as a method used for establishing core operating limits. The analyses using ASTRUM demonstrated that the applicable acceptance criteria in 10 CFR 50.46 are met. Margins of safety for large break LOCAs include quantitative limits for fuel performance established in 10 CFR 50.46. These acceptance criteria are not being changed by this proposed new methodology. Large break LOCA analyses performed consistent with the methodology in NRC approved WCAP-16009-P-A, including applicable assumptions, limitations and conditions, demonstrate that 10 CFR 50.46 acceptance criteria are met; thus, this change does not involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4.3 Conclusion Since the issuance of 10 CFR 50, Appendix K, the NRC and the nuclear industry have developed improved thermal-hydraulic computer codes and modes that- more accurately and realistically perform accident analysis calculations. Westinghouse has developed the ASTRUM methodology for performing best estimate LBLOCA analyses as documented in WCAP-16009-P-A. The NRC has approved WCAP-16009-P-A for application to Westinghouse four loop plants. WCGS is a Westinghouse four-loop plant.

A LBLOCA analysis has been performed using the ASTRUM methodology. The results demonstrate that the acceptance criteria of 10 CFR 50.46 are met. The proposed change incorporates the best estimate LBLOCA analysis using ASTRUM into the WCGS licensing basis and revises TS Section 5.6.5b. to add WCAP-16009-P-A to the list of NRC-approved methods for establishing core operating limits.

Based on the considerations discussed above, 1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 2) such activities will be conducted in compliance with the Commission's regulations, and 3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Attachment I to ET 10-0025 Page 30 of 30

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Revision 0, January 2005.

2. Letter from H. N. Berkow (USNRC) to J. A. Gresham (Westinghouse Electric Company),

"Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" (TAC NO. MB9483)," November 5, 2004.

3. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.

4. Federal Register, Volume 39, Number 3, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled Nuclear Power Reactors," January 4, 1974.

5. SECY-83-472, "Emergency Core Cooling System Analysis Methods," November 17, 1983.

6. Regulatory Guide 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance," May 1989.

7. NUREG/CR-5249, "Quantifying Reactor Safety Margins: Application of Code Scaling Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant Accident," December 1989.

8. License Amendment No. 144, "Wolf Creek Generating Station - Issuance of Amendment RE: Relocation of Cycle Specific Parameters to the Core Operating Limits Report (TAC NO. MB1638)," March 28, 2002.

Attachment II to ET 10-0025 Page 1 of 2 Proposed Technical Specification Changes (Mark-up)

Attachment II to ET 10-0025 Page 2 of 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

4. WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control -

FQ Surveillance Technical Specification."

5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."

6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7."

valuatio de ing the SH Code

8. WCAP-1-1596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Coa W, System

9. Design 'W12065-P-,A, "ANC A98 W ouse Adacd o ftestingh Compter o'ýde.for Pressurized Water Reactor Cores."

9. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code."

10. WCAP-1 2610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."

11. WCAP-8745-P-A, "Design Bases for the Thermal Power AT and Thermal Overtemperature AT Trip Functions."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

LUJCAF- %Iooq -P-A, "eiihc.LS 2skLOAEve~1Xk#.IozE; -Ae+V%0A6I05l LAtsn.5 Kr A vtomx~e Uv~ v%

LUry W\"4.-iko&(A (continued)

Wolf Creek- Unit 1 5.0-26 Amendment No. 123, 1*2, 144, 158, 1-64, 179

Attachment III to ET 10-0025 Page 1 of 2 Revised Technical Specification Pages

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

4. WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control -

FQ Surveillance Technical Specification."

5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."

6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7."

7. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."

8. WCAP-1 1596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."

9. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code."

10. WCAP-1 2610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."

11. WCAP-8745-P-A, "Design Bases for the Thermal Power AT and Thermal Overtemperature AT Trip Functions."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

Wolf Creek - Unit 1 5.0-26 Amendment No. 123, 142, 144, 158, 464,--,79,

Attachment IV to ET 10-0025 Page 1 of 3 Proposed TS Bases Changes (for information only)

Attachment IV to ET 10-0025 Page 2 of 3 Accumulators B 3.5.1 BASES APPLICABLE The worst case small break LOCA analyses also assume a time delay SAFETY ANALYSES before pumped flow reaches the core. For the larger range of small (continued) breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated primarily by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and ECCS pumps play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following a LOCA:

a. Maximum fuel element cladding temperature is < 2200°F;

b. Maximum cladding oxidation is < 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is

< 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; and

d. Core is maintained in a coolable geometry.

Since the accumulators empty themselves by the beginning stages of the reflood phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

Forodth the* rg ansal a* n alyspe, a n ifnal cqfa*

+/-.,uLarumulr ~d.

• • wd~o The contained water volume is the same as the available deliverable volume for the accumulators. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and V~~tYuml*6 4ke waeolm useA" 'e tore iWnAg tOAo~~

Wolf Creek - Unit 1 B 3.5.1-3 Revision 0

Attachment IV to ET 10-0025 Page 3 of 3 Accumulators B 3.5.1 BASES APPLICABLE The minimum boron concentration limit is used in the post LOCA boron SAFETY ANALYSES concentration calculation. The calculation is performed to assure reactor (continued) subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump boron concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH.

The small break LOCA performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator th; Loxr,-,k LoCA integrity.

acr-u v-, 's.'v -*0 I The effects on containment mass and energy releases from the

• Via m -of sts.kaccumulators are accounted for in the appropriate analyses (Refs. 1 and 3).

The accumulators satisfy Criterion 2 and Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated.

For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, Wolf Creek - Unit 1 B 3.5.1-4 Revision 0

Attachment V to ET 10-0025 Page 1 of 3 Proposed COLR Changes (for information only)

Attachment V to ET 10-0025 Page 2 of 3 Wolf Creek Generating Station WCREEK 'NUCLEAR OPERATING CORPORATION Cycle 18 Core Operating Limits Report Revision 0 fte*'esn1qh ouse ECCS" 10266-P-A,/, dendum 2, Pision 2, "TheI,81 Versionpothe inghouse E CCS Evaluatio ,Vlodel Using thBASH Code ddendum 2:

ASH Metho ogy Improve nts and Relia ity Enhance .nts," May 19 NRC lette dated Janua 0, 1988, "Acc tance for R rencing Topi Repo ddendum 2 t CAP- 10266, evision 2, SH Methodol y I.Imp vements and eliability Enha ements."

8. WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988.

NRC Safety Evaluation Report dated May 17, 1988, "Acceptance for Referencing of Westinghouse Topical Report WCAP-1 1596 - Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."

9. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"

September 1988.

NRC letter dated June 23, 1986, "Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP."

10. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.

NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance for Referencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel Assembly Reference Core Report' (TAC NO. 77258)."

NRC Safety Evaluation Report dated September 15, 1994, "Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1,

'Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC NO. M86416)."

11. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Function." September 1986.

NRC Safety Evaluation Report dated April 17, 1986, "Acceptance for Referencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions."'

Page 15 of 15

Attachment V to ET 10-0025 Page 3 of 3 INSERT A WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Revision 0, January 2005.

NRC letter dated November 5, 2004, "Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" (TAC NO. MB9483)."