ML102980447

Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR Best-Estimate Large Break Loss of Coolant Accident (BE-LBLOCA)

ML102980447
Person / Time
Site:	North Anna
Issue date:	10/21/2010
From:	Hartz L Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
10-575
Download: ML102980447 (69)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 100FR5O.90 October 21, 2010 U. S. Nuclear Regulatory Commission Serial No.10-575 ATTN: Document Control Desk NL&OS/ETS RO Washington, D. C. 20555 Docket Nos. 50-338/339 License Nos. NPR-4/7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST(LAR)

ADDITION OF ANALYTICAL METHODOLOGY TO COLR BEST-ESTIMATE LARGE BREAK LOSS OF COOLANT ACCIDENT (BE-LBLOCA)

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers NPR-4 and NPR-7 for North Anna Power Station Units 1 and 2, respectively. The proposed LAR requests the inclusion of the Westinghouse BE-LBLOCA analysis methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis of LBLOCA to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) in Technical Specification (TS) 5.6.5.b. This LAR also removes four obsolete COLR references that supported North Anna Improved Fuel (NAIF) product (i.e., Westinghouse Vantage 5H).

The NAIF product is not planned to be used in future North Anna cores.

A BE-LBLOCA analysis using ASTRUM has been completed for North Anna Units 1 and 2 and provides the technical basis for the NRC's review and approval of the implementation of the Westinghouse BE-LBLOCA using ASTRUM for the North Anna analysis of a LBLOCA. The North Anna analysis was performed in compliance with the NRC conditions and limitations identified in WCAP-16009-P-A (Reference 1). The analysis employed a plant-specific adaptation of the ASTRUM evaluation model, consisting of increasing the number of circumferential noding stacks in the downcomer region from 3 to 9. This modeling is further described in Attachment 1. Based on the analysis results, it is concluded that North Anna Units 1 and 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

A discussion of the proposed changes and the technical basis for the NRC's review and approval of the implementation of the Westinghouse BE-LBLOCA methodology using ASTRUM for the North Anna analysis of a LBLOCA is provided in Attachment 1. The marked-up and typed proposed TS pages are provided in Attachments 2 and 3, respectively. Attachment 4 provides the TS Bases changes, which are provided for information only.

We have evaluated the proposed amendment and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for our

Serial No.10-575 Docket Nos. 50-338/339 Page 2 of 3 determination is included in Attachment 1. We have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite and no significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. The basis for our determination is also included in Attachment 1. The proposed amendment has been reviewed and approved by the Facility Safety Review Committee.

Dominion is currently planning to use Westinghouse RFA-2 fuel in North Anna Units 1 and 2 commencing with North Anna Unit 1, Cycle 23 (Spring 2012) and North Anna Unit 2, Cycle 23 (Spring 2013). Therefore, Dominion requests approval of the proposed amendments by November 1, 2011. Dominion also requests a 60-day implementation period following NRC approval of the requested license amendments.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Sincerely, Leslie N. Hartz Vice President - Nuclear Support Services COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Support Services of Virginia Electric and Power Company. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of her knowledge and belief.

Acknowledged before me this _j____ day of OLCVbeyK , 2010.

My Commission Expires: q 30 /)

Notary Public - \OtS COYY'fl"-I*-)C)ytd 4

INotary GINGERt LYNNPublic

• 310847 MELTON Commonwealth of Virginiak MyComaissio Expirs Apr 30, 2013 A

Serial No.10-575 Docket Nos. 50-338/339 Page 3 of 3 Attachments:

1. Discussion of Change

2. Proposed Technical Specifications Pages (Mark-Up)

3. Proposed Technical Specifications Pages (Typed)

4. Technical Specification Bases Changes (Information Only)

Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Blvd.

Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health James Madison Building - 7 th Floor 109 Governor Street Room 730 Richmond, Virginia 23219

Serial No.10-575 Docket Nos. 50-338/339 ATTACHMENT 1 DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

DISCUSSION OF CHANGE Table of Contents

2.0 BACKGROUND

3.0 PROPOSED TECHNICAL SPECIFICATIONS CHANGE 3.1 TECHNICAL SPECIFICATIONS 5.6.5.8 CHANGE 3.2 TECHNICAL SPECIFICATIONS BASES CHANGE

4.0 TECHNICAL EVALUATION

4.1 METHOD OF THERMAL ANALYSIS FOR NORTH ANNA UNITS 1 AND 2

4.2 DESCRIPTION

OF A LARGE BREAK LOCA TRANSIENT 4.3 ASTRUM ANALYSIS RESULTS FOR NORTH ANNA UNITS 1 AND 2 4.4 10 CFR 50.46 REQUIREMENTS FOR NORTH ANNA UNITS 1 AND 2 4.5 EVALUATION OF PRE-TRANSIENT AND TRANSIENT OXIDATION 5.0 SAFETY SIGNIFICANCE

6.0 REGULATORY EVALUATION

6.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 6.2 DETERMINATION OF No SIGNIFICANT HAZARDS CONSIDERATION 6.3 ENVIRONMENTAL ASSESSMENT 6.4 REGULATORY CONCLUSION

7.0 CONCLUSION

S

8.0 REFERENCES

Page 1 of 50

DISCUSSION OF CHANGE

1.0 INTRODUCTION

Virginia Electric and Power Company (Dominion) proposes this license amendment request (LAR) for Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed change incorporates the Westinghouse Best-Estimate Large Break Loss of Coolant Accident (BE-LBLOCA) analysis methodology using the Automated Statistical TReatment of Uncertainty Method (ASTRUM) for the analysis of large break loss of coolant accident (LBLOCA) to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) in Technical Specification (TS) 5.6.5.b.

A BE-LBLOCA analysis using ASTRUM has been completed for North Anna Units 1 and 2.

Section 4.0 provides the technical basis for the USNRC review and approval of the implementation of the Westinghouse BE-LBLOCA using ASTRUM for the North Anna analysis of the LBLOCA event. The North Anna analysis was performed in compliance with the NRC conditions and limitations identified in WCAP-16009-P-A (Reference 1). Based on the analysis results, it is concluded that North Anna Units 1 and 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

This LAR also removes four COLR references that support North Anna Improved Fuel (NAIF) product (i.e., Westinghouse Vantage 5H). The NAIF product is not planned to be used in future North Anna cores. The four references being removed are the Appendix K LBLOCA methods (i.e., BASH and BART A-i) (References 3, 4 and 5) and the COBRA/WRB-1 thermal-hydraulic code (Reference 6).

Theproposed LAR has been reviewed, and it has been determined that no significant hazards consideration exists as defined in 10 CFR 50.92. In addition, it has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed Technical Specification change.

2.0 BACKGROUND

Dominion plans to purchase fuel assemblies from Westinghouse for use at North Anna Power Station Units 1 and 2. These assemblies are planned to be inserted in Units 1 and 2, commencing with Cycle 23 for both units (Spring 2012 - Unit 1 and Spring 2013 - Unit 2).

The Westinghouse 17x17 RFA-2 fuel product is a replacement for the resident fuel product, which is the AREVA Advanced Mark-BW (AMBW). The Westinghouse RFA-2 fuel product contains modified mid-grids and modified intermediate flow mixer grids (IFMs). Dominion also plans to use Optimized ZIRLOTM with the RFA-2 fuel product. A separate exemption request for this was submitted to the NRC on May 6, 2010 (Reference 13) and is currently under review.

Best-estimate loss of coolant accident analyses have been completed for the North Anna Power Station Units 1 and 2. This LAR for North Anna Units 1 and 2 requests approval to Page 2 of 50

apply the Westinghouse BE-LBLOCA analysis methodology using ASTRUM. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A (Reference 1). Additionally, a plant-specific adaptation of the Westinghouse best-estimate analysis was used for the North Anna Units 1 and 2 analyses. Further discussion of the plant-specific adaptation is provided in the "Method of Thermal Analysis for North Anna Units 1 and 2" section (Section 4.1).

This report summarizes the application of the Westinghouse ASTRUM BE-LBLOCA evaluation model to North Anna Power Station Units 1 and 2 for the large break LOCA accident analysis. Both North Anna 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 BE-LBLOCA analysis impact assessments.

3.0 PROPOSED TECHNICAL SPECIFICATIONS CHANGE 3.1 Technical Specifications 5.6.5.b Change The current North Anna TS 5.6.5.b, Core Operating Limits Report (COLR), contains references to the analytical methods used to determine the core operating limits. The specific proposed changes are identified below and mark-ups are included in Attachment 2 and 3. The-proposed additional method supports core limits, TS 5.6.5.a.7 "Heat Flux Hot Channel Factors" and TS 5.6.5.a.8 "Nuclear Enthalpy Rise Hot Channel Factor."

TS 5.6.5.b, CORE OPERATING LIMITS REPORT (COLR)

TS 5.6.5.b is revised to delete the current TS COLR Methodologies 2, 3,_ 4 and 9, to add a new reference that reflects ASTRUM (Reference 1), and to renumber TS COLR Methodologies 5-8 and 10-19 to 3-16. The new TS 5.6.5.b reference is as follows:

2. Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," as approved by NRC Safety Evaluation Report dated xx/xx/xx.

3.2 Technical Specifications Bases Change The TS Bases changes are made to differentiate the usage of accumulator water level (volume) between SBLOCA and LBLOCA methodologies. The SBLOCA methodologies use a nominal contained accumulator water volume. Whereas, the best-estimate or realistic LBLOCA methodologies sample the accumulator water volume over a given range. The other bases changes are to replace "10 CFR 50, Appendix K" with "10 CFR 50.46". The use of 10 CFR 50.46 allows for the use of both best-estimate LBLOCA and deterministic small break LOCA methodologies, since it invokes the requirements of Appendix K. The TS Bases markups are provided for information in Attachment 4.

Page 3 of 50

4.0 TECHNICAL EVALUATION

4.1 Method of Thermal Analysis for North Anna Units I and 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 (Reference 2), 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 cladding 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:

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 LBLOCA 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 8). 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 1998, 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 LBLOCA. This decision was based on an improved understanding of LBLOCA 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 LBLOCA analysis, an assessment of the uncertainty of the best-estimate calculations. It further requires that this analysis uncertainty be included when comparing Page 4 of 50

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 9).

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 (Reference 10). This method outlined an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LBLOCA analysis.

A LBLOCA 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 (Reference 7).

More recently, Westinghouse developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical TReatment of Uncertainty Method (Reference 1). This method is still based on the Code Qualification Document (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 where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing applications in WCAP-16009-P-A (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.

The North Anna Unit 1 and 2 analyses are in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A (Reference 1) as applicable to the ASTRUM methodology. Section 13-3 of WCAP-1 6009-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 Report appended to this WCAP.

The North Anna Units 1 and 2 BE-LBLOCA analyses.were performed using a more refined downcomer noding as was previously used for D.C. Cook Unit 1 (References 11 and 12).

Preliminary results for D.C. Cook Unit 1 with the as-approved ASTRUM method were observed to predict non-physical behaviors which were attributed to overly conservative aspects of the model. Consequently, an adaptation of ASTRUM was developed to better model the downcomer region by increasing the number of circumferential noding stacks by a factor of three. For North Anna Units 1 and 2, this increases the number of downcomer stacks modeled from three (one per cold leg) to nine (three per cold leg). The detailed radial noding of the vessel wall remains unchanged from the approved ASTRUM Evaluation Model and therefore does not change the historically approved method for addressing downcomer boiling during reflood. This finer nodalization has been assessed against experimental data, as described in Reference 11. The qualification of the Page 5 of 50

increased downcomer noding contained in Reference 11 has been reviewed specifically for application to North Anna Units 1 and 2. It was determined that the validation conclusions are valid and that the downcomer modeling is suitable for use as a part of the ASTRUM evaluation model for North Anna Units 1 and 2. The NRC Safety Evaluation Report for the D.C. Cook Unit 1 plant-specific application for the Large Break LOCA analysis is provided in Reference 12.

4.2 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. Traditionally, cold leg breaks have been limiting 'for large break LOCA. This location is the one where flow stagnation in the, core appears most likely to occur. Scoping studies with WCOBRA/TRAC have confirmed that the cold leg remains the limiting break location (Reference 7).

Immediately following the cold leg break, a rapid RCS 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 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 repressurization. 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.

4.3 ASTRUM Analysis Results for North Anna Units I and 2 Table 1 lists the major plant parameter assumptions used in the BE-LBLOCA analysis for the North Anna Power Station Units 1 and 2. Due to the differences in plant configuration, Page 6 of 50

namely that North Anna Unit 1 has an upflow barrel/baffle configuration and North Anna Unit 2 has a downflow configuration, two separate plant models were made and two separate analyses were completed (one for each Unit). The results of the North Anna Unit 1 ASTRUM analysis are summarized in Table 2, and the results of the North Anna Unit 2 ASTRUM analysis are summarized in Table 3. Tables 4 and 5 contain the sequence of events for the limiting PCT transients for North Anna Unit 1 and Unit 2, respectively.

The effects of Optimized ZIRLOTM cladding on the BE-LBLOCA analyses described herein have been considered for North Anna Units 1 and 2. It has been concluded that the LOCA ZIRLO models are acceptable for application to Optimized ZIRLOTM cladding in ECCS performance analyses. Therefore, the use of Optimized ZIRLOTM cladding is deemed acceptable for North Anna Units 1 and 2. No PCT penalty will be required for the North Anna Units 1 and 2 BE-LBLOCA analyses with 17x17 RFA-2 Fuel when Optimized ZIRLOTM is implemented.

The scatter plots presented in Figure 1 (Unit 1) and Figure 16 (Unit 2) show the effect of the effective break area on the analysis PCT for North Anna Units 1 and 2, respectively.

The effective break area is calculated'by multiplying the discharge coefficient (CD) by the sampled value of the break area, normalized to the cold-leg cross sectional area. Figures 1 and 16 are provided to show that the break area is a significant contributor to the variation in PCT.

From the 124 calculations performed as part of the ASTRUM analysis, a limiting case is determined for each Unit. For Unit 1, one case proved to be noticeably limiting for PCT, while also being one of the Top Five limiting cases for 'LMO and CWO. For Unit 2, the same case proved to be the limiting PCT, LMO, and CWO transient. Figure 2 (Unit 1) and Figure 17 (Unit 2) show the predicted HOTSPOT cladding temperature transient at the PCT location for the limiting cases. The HOTSPOT PCT plots include local uncertainties applied to the Hot Rod. Figure 3 (Unit 1) and Figure 18 (Unit '2) present the WCOBRA/TRAC PCT transient predicted for the limiting cases. These figures do not account for any local uncertainties.

Figures 4 through 14 (Unit 1) and Figures 19 through 29 (Unit 2) illustrate the key major response parameters for the limiting PCT transient. The reference point for the lower plenum liquid level presented in Figure 9 (Unit 1) and Figure 24 (Unit 2) is the bottom of the vessel (9.8 feet below the bottom of active fuel). The reference point for the downcomer liquid level presented in Figure 14 (Unit 1) and Figure 29 (Unit 2) is the bottom of the downcomer (6.2 feet below the bottom of active fuel). The reference point for the core collapsed liquid level presented in Figure 13 (Unit 1) and Figure 28 (Unit 2) is the bottom of active fuel.

The containment backpressure utilized for the BE-LBLOCA analysis as compared to the calculated containment backpressure is provided in Figure 15 for Unit 1 and Figure 30 for Unit 2. The worst single failure for the BE-LBLOCA analysis is the loss of one train of ECCS injection (consistent with the ASTRUM Topical); however, the containment systems which would reduce containment pressure are modeled for the BE-LBLOCA containment backpressure calculation.

Page 7 of 50

4.4 10 CFR 50.46 Requirements for North Anna Units 1 and 2 It must be demonstrated that there is a high level of probability that the limits set forth in 10 CFR 50.46 are met. The demonstration that these limits are met for North Anna Unit 1 and North Anna Unit 2 is as follows:

(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 Unit 1 limiting case is 1852 0 F and the Unit 2 limiting case is 1871°F, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(1), "Peak Cladding Temperature less than 2200 0 F," is satisfied. The results are shown in Table 2 for Unit 1 and Table 3 for Unit 2.

(b)(2) The maximum cladding oxidation corresponds to a bounding estimate of the 95thr percentile LMO at the 95-percent confidence level. Since the resulting LMO for the Unit 1 limiting case is 4.67 percent and the Unit 2 limiting case is 3.53 percent, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(2),

"Local Maximum Oxidation of the cladding less than 17 percent of the total cladding thickness before oxidation", is satisfied. The results are shown in Table 2 for Unit 1 and Table 3 for Unit 2.

(b)(3) The limiting CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. The limiting Hot Assembly Rod (HAR) total maximum oxidation is 0.38 percent for Unit 1 and 0.42 percent for Unit 2. A detailed CWO calculation takes advantage of the, core power census that includes many lower power assemblies. Because there is significant margin to the regulatory limit, the CWO value can be conservatively chosen as that calculated for the limiting HAR. A detailed CWO calculation is therefore not needed because the outcome will always be less than the limiting HAR total maximum oxidation. Since the resulting CWO is 0.38 percent for Unit I and 0.42 percent for Unit 2, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(3), "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 satisfied. The results are shown in Table 2 for Unit 1 and Table 3 for Unit 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 have been satisfied for Best-Estimate LBLOCA applications. The approved methodology (Reference 1) specifies that effects of LOCA and seismic loads on core geometry do not need to be considered unless grid crushing extends beyond the 28 assemblies in the low-power channel. This situation has been calculated not to occur for North Anna Units 1 and 2. The actions, automatic or manual, that are currently in place at the North Anna Power Station Unit 1 and Unit 2 to maintain long-term cooling remain unchanged Page 8 of 50

with the application of the ASTRUM methodology (Reference 1). Therefore, acceptance criterion (b)(4) is satisfied.

(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 manual actions that are currently in place to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology (Reference 1). Therefore, acceptance criterion (b)(5) is satisfied.

Based on the ASTRUM analysis results (Table 2 and Table 3), it is concluded that North Anna Unit 1 and Unit 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

4.5 Evaluation of Pre-Transient and Transient Oxidation The pre-existing oxidation was not factored into the Local Maximum Oxidation results presented for North Anna Unit 1 and Unit 2 in Tables 2 and 3, respectively. The maximum expected total of the normal operation (pre-transient) and BE-LBLOCA transient oxidation, for any time in life, was considered for North Anna Units 1 and 2. The pre-transient oxidation increases with burnup, from zero at beginning-of-life (BOL) to a maximum value at the discharge of fuel (end-of-life, or EOL). The transient oxidation has been calculated throughout the entire life of the fuel. It has been confirmed that the sum of the pre-transient plus transient oxidation remains below 17% at all times in life for North Anna Units 1 and 2.

Page 9 of 50

Table 1: Major Plant Parameter Assumptions Used in the BELOCA Analyses for North Anna Units 1 and 2 Parameter Value Plant Physical Description

• SG Tube Plugging :5 7%

Plant Initial Operating Conditions

• Maximum Reactor Power 5 2951 MWt

• Peaking Factors FQ 5 2.32; FAH 5 1.65

• Axial Power Distribution See Figure 31 Fluid Conditions

" TAVG 580.8 - 5.7°F < TAVG 5 586.8 + 5.7 0 F

• Pressurizer Pressure 2250 - 30 psia < PRCS 5 2250 + 30 psia

" Reactor Coolant Flow 92,800 gpm per loop

• Accumulator Temperature 86°F < TACC 5 105°F

• Accumulator Pressure 590.4 psia < PACC 5 705 psia 3

• Accumulator Water Volume 950.8 ft < VACC 5 986.3 ft3

" Accumulator Boron Concentration 2400 ppm Accident Boundary Conditions

• Single Failure Assumptions Loss of one ECCS train

• Safety Injection Flow Minimum

" Safety Injection Temperature 38oF < SI Temp < 52'F

" Safety Injection Initiation Delay Time - 15 seconds (with offsite power)

< 29 seconds (with LOOP)

• Containment Pressure Bounded (minimum); See Figures 15 and 30 Page 10 of 50

Table 2: Best-Estimate Large Break LOCA Analysis Results for North Anna Unit I Results Criteria 95/95 Peak Cladding Temperature 1852°F < 2200'F 95/95 Local Maximum Oxidation 4.67% < 17%

95/95 Core Wide Oxidation 0.38% < 1%

Table 3: Best-Estimate Large Break LOCA Analysis Results for North Anna Unit 2 Results Criteria 95/95 Peak Cladding Temperature 1871°F < 2200°F 95/95 Local Maximum Oxidation 3.53% < 17%

95/95 Core Wide Oxidation 0.42% < 1%

Page 11 of 50

Table 4: North Anna Unit I Best-Estimate Large Break LOCA Sequence of Events for the Limiting PCT Case Event Time After Break (sec)

Start of Transient 0.0 Safety Injection Signal 4.9 Accumulator Injection Begins 10.0 Safety Injection Begins 19.9 End of Blowdown 24.0 Bottom of Core Recovery 31.0 Accumulator Empty(1 ) 33.3 PCT Occurs HOTSPOT PCT (Global + Local) 200 WCOBRA/TRAC PCT (Global) 204 Core Quenched 500 End of Transient 700 Note:

1. Accumulator liquid injection ends.

Page 12 of 50

Table 5: North Anna Unit 2 Best-Estimate Large Break LOCA Sequence of Events for the Limiting PCT Case Event Time After Break (sec)

Start of Transient 0.0 Safety Injection Signal 4.8 Accumulator Injection Begins 10.0 End of Blowdown 23.0 Bottom of Core Recovery 30.5 Safety Injection Begins 33.8 1

Accumulator Empty( ) 34.5 PCT Occurs HOTSPOT PCT (Global + Local) 90 WCOBRAITRAC PCT (Global) 94 Core Quenched 500 End of Transient 700 Note:

1. Accumulator liquid injection ends.

Page 13 of 50

PCT vs. (CD

• A) (All 124 Cases) 0

• PCT DEG 0 0 0 PCT DEGCL [dog F]

A A PCT-SPL 0 0 0 PCT SPLIT [dog F]

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.................. .... . U...- IN, ... A.

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,. AAA AM A 1400-C-) A.............

nA A.

A A :A A 1200- AM A AA A4. . . . .

A:............... .........

. l . . . . . . . . ..

1000" I l l s I I I I I l l I l l I a

.nn..I. . . m. I I I Up"'

0 0.5 1 1.5 2 2.5 3 CD

• Abreak/ACL Figure 1: North Anna Unit I HOTSPOT PCT versus Effective Break Area Scatter Plot (CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area)

Page 14 of 50

- TCLAD2 0 0 0 HOTSPOT CLAD TEMP 2 E!1000 Q-E 500-0- i t I I I I I I I I I It Ii I i I I I I I I I I I I I Ii 0 100 200 300 400 500 600 700 Time Ater BreOk (s)

Figure 2: North Anna Unit I Limiting PCT Case - HOTSPOT Cladding Temperature Transient Page 15 of 50

Temperature (F)

PCT 1 0 0 PEAK CLADDING TEMP.

Elevation (ft)

.-... PCT- LC 1 0 0 PEAK CLAD TEMP LOC.

1800 1600-1400 I'

1200 I I

I

.II 1 1000-1 I E I ' a)j 0 100 200 300 400 500 600 700 Time Ater Break (s)

Figure 3: North Anna Unit I Limiting PCT Case - WC/T Cladding Temperature Transient and PCT Location Page 16 of 50

RMVM 36 1 0 MASS FLOWRATE 60000-50000 40000-ct)

E 0

L- 20000 U)

U) 0 10 20 30 40 50 Time PCter Breek (S)

Figure 4: North Anna Unit 1 Limiting PCT Case - Vessel Side Break Flow Page 17 of 50

RMVM 35 4 0 MASS FLOWRATE 30000-25000-20000 E

= 15000 0

S10000 U)

U) 0 10 Time20 After Break30 (s) 40 50 Figure 5: North Anna Unit 1 Limiting PCT Case - Pump Side Break Flow Page 18 of 50

ALPN 24 1 0 INTACT LOOP ALPN 34 1 0 BROKEN LOOP I

I 0.8-I I

- I I

0. I o0 I 2 I 0 10 20 rBrea s) 50 Figure 6: North Anna Unit I Limiting PCT Case - Broken Loop and Intact Loop Void Fraction Page 19 of 50

FGM 13 11 0 VAP AXIAL MASS FLOW 6-4--

2--

E

.* 0-0 0

Ei -2 c,,

c,,

0 5 10 15 20 25 30 Time After Break (s)

Figure 7: North Anna Unit I Limiting PCT Case - Hot Assembly Vapor Flow Page 20 of 50

- PH 29 1 0 PRESSURE 50-0-

Ul)

U) cl 30 t 0 50 100 150 200 250 300 Time After Break (s)

Figure 8: North Anna Unit I Limiting PCT Case - Pressurizer Pressure Page 21 of 50

- LQ-LEVEL 1 0 0 COLLAPSED LIQ. LEVEL

-J 0-0)

C-)

0 100 200 300 400 500 600 700 Time After Break (s)

Figure 9: North Anna Unit I Limiting PCT Case - Lower Plenum Collapsed Liquid Level Page 22 of 50

VFMASS 0 0 0 VESSEL WATER MASS 160000 140000-120000-100000-E 80000-c=

(n) 0 100 200 Time 300 400 (s)

After Break 500 600 700 Figure 10: North Anna Unit I Limiting PCT Case - Vessel Fluid Mass Page 23 of 50

- RMVM 61 1 0 MASS FLOWRATE

~.1500*

0

= 00 W)

U) 0 20 40 60 80 100 Time PCter Break (s)

Figure 11: North Anna Unit I Limiting PCT Case - Intact Loop I Accumulator Flow Page 24 of 50

RMVM 15 5 0 MASS FLOWRATE 500 400-300-E c: 200-

-o Co Cn 0 20 lim Time A After Break (s) 80 100 Figure 12: North Anna Unit I Limiting PCT Case - Intact Loop I Safety Injection Flow Page 25 of 50

LQ-LEVEL 3 0 0 COLLAPSED LIQ. LEVEL 12 10-

'* 8"

= 6-

-o Q0 0

C-,'A 0 100 200 300 400 500 600 700 Time After Break (s)

Figure 13: North Anna Unit I Limiting PCT Case - Core Average Channel Collapsed Liquid Level Page 26 of 50

- LQ-LEVEL 16 0 0 COLLAPSED LIQ. LEVEL

2)

-J C',

0 C-)

0 100 200 300 400 500 600 700 Time After Break (s)

Figure 14: North Anna Unit I Limiting PCT Case - Average Downcomer Collapsed Liquid Level Page 27 of 50

NORTH ANNA UNIT 1 BACKPRESSURE STUDY Coco WCT 0

4 0

1 0 REF COCO RUN 0 WCT REF TR 40 35- S.. ..

30 0-21 I'

I' CL 0 100 200 After Break(s)

Time 300 400 500 Figure 15: North Anna Unit I - Lower Bound Containment Pressure Page 28 of 50

PCT vs. (CD

• A) (All 124 Cases)

SCPCT DEG 0 0 0 PCT DEGCL [deg F]

A PCT-SPL 0 0 0 PCT SPLIT [dog F]

LUIAI A

1800-A 1600- A AAAA:k

-AI

. ... l .i . .. .....

1400- A:

Q-)

A& A oA

.. "A..... i . . "

A4 U 1200-A A II.

1000- I I U. I I a a I Ia I OqA 0 I 1.5 CD

• Abreuk/ACL 2 2.5 i Figure 16: North Anna Unit 2 HOTSPOT PCT versus Effective Break Area Scatter Plot (CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area)

Page 29 of 50

' TCLAD2 0 0 0 HOTSPOT CLAD TEMP 2 1000 0D E

500--

0 0 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 0 100 200 300 4A0 500 600 700 PC Time ter Break (s)

Figure 17: North Anna Unit 2 Limiting PCT Case - HOTSPOT Cladding Temperature Transient Page 30 of 50

Temperature (F)

PCT 1 0 0 PEAK CLADDING TEMP.

Elevation (ft) 1

-PCT-LOC 0 0 PEAK CLAD TEMP LOC.

1800 1600-1400-1200 C

.1000- 0o 0>

E 0,),

3w0 400 500 Time After Break (s)

Figure 18: North Anna Unit 2 Limiting PCT Case - WC/T Cladding Temperature Transient and PCT Location Page 31 of 50

- RMVM 36 1 0 MASS FLOWRATE E

-* 30000' 0

=r: 20000.

C',

C',

0 10 20 30 40 50 Time After Break (s)

Figure 19: North Anna Unit 2 Limiting PCT Case - Vessel Side Break Flow Page 32 of 50

RMVM 35 4 0 MASS FLOWRATE 30000-25000-20000" E

-o n 10000 0 10 Time20 After Break30 (s) 40 50 Figure 20: North Anna Unit 2 Limiting PCT Case - Pump Side Break Flow Page 33 of 50

ALPN 24 1 0 INTACT LOOP ALPN 34 1 0 BROKEN LOOP 0

0

-o 0

0 10 rime20 Afe After Break (s) 40 50 Figure 21: North Anna Unit 2 Limiting PCT Case - Broken Loop and Intact Loop Void Fraction Page 34 of 50

- FGM 13 11 0 VAP AXIAL MASS FLOW E

0-0 0

-6

-a- I I I I I I I II I I I I I I I I I II I I I I II 0 5 10 15 20 25 30 Time After Break (s)

Figure 22: North Anna Unit 2 Limiting PCT Case - Hot Assembly Vapor Flow Page 35 of 50

- PN 29 1 0 PRESSURE 60-50-00-CL c,,

ci, 02 0 50 100 150 200 250 300 Time After Break (s)

Figure 23: North Anna Unit 2 Limiting PCT Case - Pressurizer Pressure Page 36 of 50

LQ-LEVEL 1 0 0 COLLAPSED LIQ. LEVEL 10-4) o-0 100 200 300 400 500 600 700 Time After Break (s)

Figure 24: North Anna Unit 2 Limiting PCT Case - Lower Plenum Collapsed Liquid Level Page 37 of 50

VVFMASS 0 0 0 VESSEL WATER MASS 160000 140000--

120000-100000-E 80000 0 100 200 300 400 500 600 700 Time After Break (s)

Figure 25: North Anna Unit 2 Limiting PCT Case - Vessel Fluid Mass Page 38 of 50

- RMVM 61 1 0 MASS FLOWRATE E

3C C,,

0n 0 20 40 60 80 100 Time After Break (s)

Figure 26: North Anna Unit 2 Limiting PCT Case - Intact Loop I Accumulator Flow Page 39 of 50

- RMVM 15 5 0 MASS FLOWRATE E

43 - -

= 200-CO, 100-0-

-100-0 20 40 60 80 100 TimePCter Break (s)

Figure 27: North Anna Unit 2 Limiting PCT Case - Intact Loop I Safety Injection Flow Page 40 of 50

LQ-LEVEL 3 0 0 COLLAPSED LIQ. LEVEL 12 10--

8-

-J6-(D a) a) _

C, 0 100 200 300 400 500 600 700 Time After Break (s)

Figure 28: North Anna Unit 2 Limiting PCT Case - Core Average Channel Collapsed Liquid Level Page 41 of 50

LQ-LEVEL 16 0 0 COLLAPSED LIQ. LEVEL 30 25-20

~-15

""15-

"-1 C-,

0 100 200 300 400 500 600 700 Time After Break (s)

Figure 29: North Anna Unit 2 Limiting PCT Case - Average Downcomer Collapsed Liquid Level Page 42 of 50

NORTH ANNA UNIT 2 BACKPRESSURE STUDY COCO 0 0 0 REF COCO RUN WCT 4 1 0 WCT REF TR 40 305 II  %

CL

• -5I........... "-.............. i...................

,n Figure 30: North Anna Unit 2 - Lower Bound Containment Pressure Page 43 of 50

0.50 0.45 -

0 0.40 0.30, 0.40: 0.425, 0.39 o

" 0.35 I

= 0.30 0

0.325, 0.255

- 0.25 - 0.453, 0.26 0.20 0.20 0.355, 5 0.205 0.42, 0.205 0.15 0.25 0.30 0.35 0.40 0.45 0.50 Power in Middle Third of Core (PMID)

Figure 31: North Anna Units I and 2 BE-LBLOCA Analysis Axial Power Shape Operating Space Envelope Page 44 of 50

5.0 SAFETY SIGNIFICANCE

SUMMARY

The Westinghouse BE-LBLOCA analysis methodology using the statistical treatment of uncertainties methodology, ASTRUM, has been approved by the USNRC (Reference 1).

In addition, the NRC has approved the use of ASTRUM for PWR safety analyses at several sites, including Surry and D.C. Cook. An analysis of the LBLOCA for North Anna Units 1 and 2 has been performed with the Westinghouse BE-LBLOCA analysis methodology using ASTRUM and is documented in Section 4.0 of this Discussion of Change. The analysis was performed in compliance with all the NRC conditions and limitations identified in WCAP-16009-P-A (Reference 1). Based on the analysis results, it is concluded that the North Anna Units 1 and 2 continue to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

6.0 REGULATORY EVALUATION

6.1 Applicable Regulatory Requirements/Criteria In accordance with 10 CFR 50.46, the conclusions of the new LBLOCA analysis show that North Anna Units 1 and 2 continue to maintain a margin of safety to the limits prescribed by the following criteria:

1. The calculated maximum fuel element cladding temperature (i.e., peak cladding' temperature) will not exceed 2,200 OF.

2. The calculated total oxidation of the cladding (i.e., maximum cladding oxidation) will 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 (i.e., maximum hydrogen generation) will 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. The calculated changes in core geometry are such that the core remains amenable to cooling.

5. After successful initial operation of the emergency core cooling system, the core temperature will be maintained at an acceptably low value and decay heat will be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

6.2 Determination of No Significant Hazards Consideration Virginia Electric and Power Company (Dominion) proposes this license amendment request (LAR), for Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively, pursuant to 10 CFR 50.90. The proposed change adds the Westinghouse Best-Estimate Large Break Loss of Coolant Accident (BE-LBLOCA) analysis methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) for the analysis of large break loss of coolant accident (LBLOCA) to the list of Page 45 of 50

methodologies approved for reference in the Core Operating Limits Report (COLR) in Technical Specification (TS) 5.6.5.b.

Virginia Electric and Power Company (Dominion) plans to implement the Westinghouse Best-Estimate Large Break Loss of Coolant Accident (BE-LBLOCA) analysis methodology using the Automated Statistical Treatment of Uncertainty Method (ASTRUM) to perform analyses of the large break loss of coolant accident (LBLOCA) at North Anna Units 1 and

2. The Westinghouse BE-LBLOCA analysis methodology using ASTRUM has been approved by the USNRC. An analysis of the LBLOCA for North Anna Units 1 and 2 has been performed with the Westinghouse BE-LBLOCA analysis methodology using ASTRUM. The analysis was performed in comhpliance with the NRC conditions and limitations identified in WCAP-1 6009-P-A. Based on the results, it is concluded that the North Anna Units 1 and 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

In accordance with the criteria set forth in 10 CFR 50.92, Dominion has evaluated the proposed TS change and determined that the change does not represent a significant hazards consideration. The following is provided in support of this conclusion:

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

Response: No.

No physical plant changes are being made as a result of using the Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis methodology. The proposed TS change simply involves updating the references in TS 5.6.5.b, Core Operating Limits Report (COLR), to reference the Westinghouse BE-LBLOCA analysis methodology, which is an NRC approved methodology, and to delete unnecessary references. Therefore, the probability of LOCA occurrence is not affected by the change. Further, the consequences of a LOCA are not increased, since the BE-LBLOCA analysis has demonstrated that the performance of the Emergency Core Cooling System (ECCS) continues to conform to the criteria contained in 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors." No other accident consequence is potentially affected by this change.

Systems will continue to be operated in accordance with current design requirements under the new analysis, therefore no new components or system interactions have been identified that could lead to an increase in the probability of any accident previously evaluated in the Updated Final Safety Analysis Report (UFSAR). No changes were required to the Reactor Protection System (RPS) or Engineering Safety Features (ESF) setpoints because of the new analysis methodology.

An analysis of the LBLOCA accident for North Anna Units 1 and 2 has been performed with the Westinghouse BE-LBLOCA analysis methodology using ASTRUM. The analysis was performed in compliance with the NRC conditions and Page 46 of 50

limitations as identified in WCAP-1 6009-P-A. Based on the analysis results, it is concluded that the North Anna Units 1 and 2 continue to satisfy the limits prescribed by 10 CFR 50.46.

There are no changes to assumptions of the radiological dose calculations. Hence, there is no increase the predicted radiological consequences of accidents postulated in the UFSAR.

Therefore, neither the probability of occurrence nor the consequences of an accident previously evaluated is significantly increased.

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

Response: No.

The use of the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not impact any of the applicable design criteria and pertinent licensing basis criteria continue to be met. Demonstrated adherence to the criteria in 10 CFR 50.46 precludes new challenges to components and systems that could introduce a new type of accident. Safety analysis evaluations have demonstrated that the use of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is acceptable.

Design and performance criteria continue to be met and no new single failure mechanisms have been created. The use of the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not involve any alteration to plant equipment or procedures that would introduce any<new or unique operational modes or accident precursors. Furthermore, no changes have been made to any RPS or ESF actuation setpoints. Based on this review, it is concluded that no new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

Therefore, the possibility for a new or different kind of accident from any accident previously evaluated is not created.

3. Does this change involve a siqnificant reduction in a margqin of safety?

Response: No.

It has been demonstrated that the analytical technique used in the Westinghouse BE-LBLOCA analysis methodology using ASTRUM realistically describes the expected behavior of the reactor system during a postulated LOCA. Uncertainties have been accounted for as required by 10 CFR 50.46. A sufficient number of LOCAs with different break sizes, different locations, and other variations in properties have been considered to provide assurance that the most severe postulated LOCAs have been evaluated. The analysis has demonstrated that the acceptance criteria contained in 10 CFR 50.46 continue to be satisfied.

Page 47 of 50

Therefore, it is concluded that this change does not involve a significant reduction in the margin of safety.

Based on the above information, Dominion concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

6.3 Environmental Assessment The proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(1) as follows:

(i) The proposed change involves no significant hazards consideration.

As described in Section 6.2 above, the proposed change involves no significant hazards consideration.

(ii) There are no significant changes in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change does not involve the installation of any new equipment or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite. Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

.i The proposed change does not involve physical plant changes or introduce any new modes of plant operation. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion concludes that, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.4 Regulatory Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by implementation of the proposed TS change, (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.

Page 48 of 50

7.0 CONCLUSION

S The use of the Westinghouse BE-LBLOCA analysis methodology with ASTRUM has been demonstrated to provide acceptable results for the analysis of the LBLOCA at North Anna Power Station Units 1 and 2. The Westinghouse BE-LBLOCA analysis methodology using ASTRUM will be added to the list of methodologies approved for reference in the COLR in TS 5.6.5.b upon USNRC approval. The references to the Westinghouse Appendix K methods for LBLOCA and the COBRA thermal-hydraulic code, which only support the NAIF product, will be simultaneously removed from the list of methodologies approved for reference in the COLR in TS 5.6.5.b upon USNRC approval.

Page 49 of 50

8.0 REFERENCES

1. 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).

2. "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.

3. WCAP-9220-P-A, "Westinghouse ECCS Evaluation Model-I1981 Version," February 1982.

4. WCAP-9561-P-A, "BART A-I: A Computer Code for The Best Estimate Analysis of Reflood Transients-Special Report: Thimble Modeling In W ECCS Evaluation Model."

5. WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.

6. VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," July 1990.

7. 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).

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

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

Regulatory Guide 1.157, USNRC, May,1989.

10.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.

11. "Donald C. Cook, Unit 1 - License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," ADAMS Accession Number ML080090268, December 27, 2007.

12. "Donald C. Cook Nuclear Plant Unit 1 - Issuance of Amendment to Renewed Facility Operating License Regarding use of the Westinghouse ASTRUM Large Break Loss-of-Coolant Accident Analysis Methodology," TAC MD7556, ADAMS Accession Number ML082670351, October 17, 2008.

13.Letter to Document Control Desk (USNRC) from J. A. Price (Dominion), "North Anna Power Station Units 1 and 2, Proposed License Amendments and Exemption Request for use of Optimized ZIRLO Fuel Rod Cladding," date May 6, 2010 [ADAMS Accession ML101260517].

Page 50 of 50

Serial No.10-575 Docket Nos. 50-338/339 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS PAGES (MARK-UP)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

2. Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large Break LOCA ng Requirements 5.6 Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method/

(ASTRUM)," as approved by NRC Safety Evaluation Report dated xx/xx/xx.

5,6 Reporrtir qg Requirements

.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. VEP-FRD-42-A, "Reload Nuclear Design Methodology."

2. WCAP 9220 P A, "WESTINGHOUSE EGGS EVALUATIN M)DEL 1981 VERSION."

3. WCAP 9561 P A, "BART A I, A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS SPECIAL REPORT:

IIIJLE MODELING IN W [CC [VALUATION KODEL."

4. WCAP 10266 P A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASHl Code."

. WCAP-10054-P-A, "Westinghouse Small Break.ECCS Evaluation Model Using the NOTRUMP Code."

n4 --6-. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code."

I--7-. WCAP-12610, "VANTAGE+ FUEL ASSEMBLY-REFERENCE CORE REPORT."

FI6--& VEP-NE-2-A, "Statistical DNBR Evaluation Methodology."

9. VEP NE 3 A, "Qualif.cati en ef the WRB 1 CI" Ccrrelatien in the Virginia Power COBRA Cmdrp 1

[*l-0-O. VEP-NE-1-A, "VEPCO Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications."

EZ-1-. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function."

Fl-l-2, WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report."

10.-1-3-. BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel."

(continued)

North Anna Units 1 and 2 5.6-3 Amendments 239/220

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

b. (continued)

.- i-4-. BAW-10199P-A, "The BWU Critical Heat Flux Correlations."

5-.

BAW-10170P-A, "Statistical Core Design for Mixing Vane Cores."

13. ] . EMF-2103 (P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

14. --. EMF-96-029 (P)(A), "Reactor Analysis System for PWRs."

F5* -- 8-. BAW-10168P-A, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"

Volume II only (SBLOCA models).

-6.4-9-. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including'Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code."

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.

5.6.6 PAM Report When a report is required by Condition B of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation,",a report shall be submitted within the following 14 days. The report shall outline the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

North Anna Units 1 and 2 5..6-4 Amendments 248/228

Serial No.10-575 Docket Nos. 50-338/339 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS PAGES (TYPED)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

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

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. VEP-FRD-42-A, "Reload Nuclear Design Methodology."

2. Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," as approved by NRC Safety Evaluation Report dated xx/xx/xx.

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code."

4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code."

5. WCAP-12610, "VANTAGE+ FUEL ASSEMBLY-REFERENCE CORE REPORT."

6. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology."

7. WCAP-NE-1-A, "VEPCO Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications."

8. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function."

9. WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report."

10. BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel."

11. BAW-10199P-A, "The BWU Critical Heat Flux Correlations."

12. BAW-10170P-A, "Statistical Core Design for Mixing Vane Cores."

(continued)

North Anna Units 1 and 2 5.6-3 Amendments

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

b. (continued)

13. EMF-2103 (P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

14. EMF-96-029 (P)(A), "Reactor Analysis System for PWRs."

15. BAW-10168P-A, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"

Volume II only (SBLOCA models).

16. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code."

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.

5.6.6 PAM Report When a report is required by Condition B of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

North Anna Units 1 and 2 5.6-4 Amendments

Serial No.10-575 Docket Nos. 50-338/339 ATTACHMENT 4 TECHNICAL SPECIFICATION BASES CHANGES (INFORMATION ONLY)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Accumulators B 3.5.1 BASES APPLICABLE Head Safety Injection (HHSI) pumps both play a part in SAFETY ANALYSES terminating the rise in clad temperature. As break size (continued) continues to decrease, the role of the accumulators continues to decrease until they are not required and the HHSI 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-.

for small breaks, and there must be a high level of probability that the peak cladding temperature does not exceed 2200'F for large breaks;

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 discharge during the blowdown phase of a LBLOCA, they do not contribute to the long term'cooling requirements of 10 CFR 50.46. analysis while the large Fh the e small break LOCA alyses, a nominal break LOCA contained accumulator water voluIIIe itst1. For small analysis samples breaks, the accumulator water volume only affects the mass flow rate of water into the RCS since the tanks do not empty the accumulator for most break sizes analyzed. The assumed water volume has water volume over an insignificant effect upon the peak clad temperature. For a given range. large breaks, an increase in water volume canbe either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The safety analysis supports operation with a contained water volume of between 7580 gallons and 7756 gallons per accumulator.

(continued)

North Anna Units 1 and 2 B 3.5. 1-3 Revision 13

Accumulators B 3.5.1 BASES APPLICABLE The minimum boron concentration setpoint is used in the post SAFETY ANALYSES LOCA boron concentration calculation. The calculation is (continued) performed to assure reactor 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 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.

lanalysis is The largeand small break LOCA peak clad temperature Dperformed 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 accumultr intime9V49.

Th* I *rn* hr~~k LOCA analysis The effects on containment mass and energy releases from the samples the accumulators are accounted for in the appropriate analyses (Ref. 1). The large break LOCA containment analyses assume accumulator that the accumulator nitrogen is discharged into the pressure over a containment, which affects transient subatmospheric given range. pressure.

The accumulators satisfy 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. Three accumulators are required to ensure that 100% of the contents of two of the accumulators will reach the core during a large break LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than two accumulators are injected during the blowdown phase of a large break LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated.

(continued)

North Anna Units 1 and 2 B 3.5.1-4 Revision 9

Containment Pressure B 3.6.4 BASES APPLICABLE one train of the Recirculation Spray System becoming SAFETY ANALYSES inoperable). The containment analysis for the DBA (Ref. 1)

(continued) shows that the maximum peak containment pressure results from the limiting design basis SLB.

The maximum design internal pressure for the containment is 45.0 psig. The LOCA and SLB analyses establish the limits for the containment air partial pressure operating range.

The initial conditions used in the containment design basis LOCA analyses were an air partial pressure of 12.3 psia and an air temperature of 115'F. This resulted in a maximum peak containment internal pressure of 42.7 psig, which is less than the maximum design internal pressure for the containment. The SLB analysis resulted in a maximum peak containment internal pressure of 43.0 psig, which is less than the maximum design internal pressure for the containment.

The containment was also designed for an external pressure load of 9.2 psid (i.e., a design minimum pressure of 5.5 psia). The inadvertent actuation of the QS System was analyzed to determine the reduction in containment pressure (Ref. 1). The initial conditions used in the analysis were 10.3 psia and 115'F. This resulted in a minimum pressure inside containment of 8.6 psia, which is considerably above the design minimum of 5.5 psia.

Controlling containment air partial pressure limits within prescribed limits ensures adequate NPSH for the recirculation spray and low head safety injection pumps following a DBA. The minimum containment air partial pressure is an initial condition for the NPSH analyses.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood 10 CFR 50.46 phase of a LOCA analysis increases with increasing containment backpressure. For the reflood phase calculations, the containment backpressure is calculated in manner designed to conservatively minimize, rather than ma i ize, the containment pressure response in accordance with l CFR 50, Appendix K (Ref. 2).

The radiological consequences analysis demonstrates acceptable results provided the containment pressure decreases to 2.0 psig in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and does not exceed 2.0 psig (continued)

North Anna Units 1 and 2 B 3.6.4-2 Revision 31

Containment Pressure B 3.6.4 BASES B.1 and B.2 If containment air partial pressure cannot be restored to within limits within the required Completion'Time,.the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment air partial pressure is within limits ensures that operation remains within the limits assumed in the containment analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was developed considering operating experience related to trending of containment pressure variations and pressure instrument drift during the applicable MODES.

Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition.

REFERENCES 1. UFSAR, Section 6.2.

^ .10CFýR50.ý46

2. 10 CPR. 50, Appendix

3. UFSAR, Section 15.4.1.7.

North Anna Units 1 and 2 B 3.6.4-4 Revision 31

QS System B 3.6.6 BASES APPLICABLE exceeded the containment design temperature was short enough SAFETY ANALYSES that there would be no adverse effect on equipment inside (continued) containment assumed to mitigate the consequences of the DBA.

Therefore, it is concluded that the calculated transient containment atmosphere temperatures are acceptable for the SLB.

The modeled QS System actuation from the containment analysis is based upon a response time associated with exceeding the containment High-High pressure signal setpoint to achieving full flow through the spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The QS System total response time of 70 seconds after Containment Pressure-High High comprises the signal delay, diesel generator startup time, and system startup time, including pipe fill time.

For certain aspects of accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure.

For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordan10 CFR 50, Appendix K (Ref. 3).

110FR 50.46 Inadvertent actuation of the QS System is evaluated in the analysis, and the resultant reduction in containment pressure is calculated. The maximum calculated reduction in containment pressure results in containment pressures within the design containment minimum pressure.

The radiological consequences analysis demonstrates acceptable results provided the containment pressure decreases to 2.0 psig in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis Accident (Ref. 4). Beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the containment pressure is assumed to be less than 0.0 psig, terminating leakage from containment.

The QS System satisfies Criterion 3 of 10 CFR 50.36(c) (2)(ii).

North Anna Units 1 and 2 B 3.6.6-3 Revision 31.

QS System B 3.6.6 BASES SURVEILLANCE SR 3.6.6.3 and SR 3.6.6.4 (continued)

REQUIREMENTS Surveillances under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power.

Operating experience has shown that these components usually pass the Surveillances when performed at an 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.6.5 With the quench spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections or an inspection of the nozzles can be performed. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. Due to the passive nature .of the design of the nozzle and the non-corrosive design of the system, a test performed following maintenance which could result in nozzle blockage is considered adequate to detect obstruction of the nozzles.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.49.

3. 10 CR 50, Appendix K.50.46

4. UFSAR, Section 15.4.1.7.

5. ASME Code for Operation and Maintenance of Nuclear Power Plants.

North Anna Units 1 and 2 B 3.6.61-6 Revision 31

RS System B 3.6.7 BASES APPLICABLE capability and the variation of containment pressure are SAFETY ANALYSES functions of service water temperature, RWST water (continued) temperature, and the containment air temperature.

The DBA analyses show that the maximum peak containment pressure of 43.0 psig results from the SLB analysis and is calculated to be less than the containment design pressure.

The maximum 309'F peak containment atmosphere temperature results from the SLB analysis and is calculated to exceed the containment design temperature for a relatively short period of time during the transient. The basis of the containment design temperature, however, is to ensure OPERABILITY of safety related equipment inside containment (Ref. 2).

Thermal analyses show that the time interval during which the containment atmosphere temperature exceeds the containment design temperature is short enough that there would be no adverse effect on equipment inside containment.

Therefore, it is concluded that the calculated transient containment atmosphere temperatures are acceptable for the SLB and LOCA.

The RS System actuation model from the containment analysis is based upon a response associated with exceeding the Containment Pressure-High High signal setpoint and RWST level decreasing below the RWST Level-Low setpoint. The containment analysis models account conservatively for instrument uncertainty for the Containment Pressure-High High setpoint and the RWST Level-Low setpoint. The RS System's total response time is determined by the time to satisfy the coincidence logic, the timer delay for the inside RS pumps, pump startup time, and piping fill time.

For certain aspects of accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure.

For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordan 10 CPR 50, Appendix K (Ref. 3).

10 CFR 50.46 e radiological consequences analysis demonstrates acceptable results provided the containment pressure decreases to 2.0 psig in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and does not exceed,2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis (continued)

North Anna Units 1 and 2 B 3.6.7-4 Revision 31

RS System B 3.6.7 BASES SURVEILLANCE SR 3.6.7.6 REQUIREMENTS (continued) These SRs ensure that each automatic valve actuates and that the casing cooling pumps start upon receipt of an actual or simulated High-High containment pressure signal. The RS pumps are verified to start with an actual or simulated RWST Level-Low signal coincident with a Containment Pressure-High High signal. The start delay times for the inside RS pumps are also verified. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was considered to be acceptable from a reliability standpoint.

SR 3.6.7.7 Periodic'inspections of the containment sump components ensure that they are unrestricted and stay in proper operating condition. The 18 month Frequency is based on the need to 'perform this Surveillance under the conditions that apply during a unit, outage and on the need to have access to the location. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience.

SR 3.6.7.8 This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment will meet its design bases objective. Either an inspection of the nozzles or an air or smoke test is performed through each spray header. Due to the passive design of the spray header and its normally dry state, a test performed following maintenance which could result in nozzle blockage is considered adequate for detecting obstruction of the nozzles.

REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.49.

3. 10 CPR 50, Appendix K. 4.D0F50.46 North Anna Units 1 and 2 B 3.6.7-9 Revision 31