ML062850195

Proposed Technical Specification Change and Supporting Safety Analyses Revisions to Address Generic Safety Issue 191

ML062850195
Person / Time
Site:	North Anna
Issue date:	10/03/2006
From:	Gerald Bichof Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-849, GSI-191
Download: ML062850195 (179)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 October 3, 2006 U. S. Nuclear Regulatory Commission Serial No.06-849 Attention: Document Control Desk NL&OS/ETS RO One White Flint North Docket Nos. 50-338/339 11555 Rockville Pike License Nos. NPF-4/7 Rockville, MVD 20852-2738 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 Pursuant to 10 CFR 50.90, Dominion hereby requests an amendment to Operating License Numbers NPF-4 and NPF-7 in the form of changes to the Technical Specifications (TS) for North Anna Power Station Units 1 and 2, respectively. The proposed change is being submitted as part of Dominion's resolution to NRC Generic Safety Issue 191 (GSI-191). In a letter dated September 1, 2005 (Serial No.05-212),

Dominion identified actions required to resolve GSI-191 and NRC Generic Letter (GL) 2004-02 for North Anna. In that letter, Dominion committed to provide this submittal in-February 2006. In a subsequent phone conversation on December 14, 2005, Dominion advised the NRC that additional time was needed to complete the North Anna specific analysis and that the associated Technical Specification change and containment analysis for North Anna would be submitted by fall of 2006.

The actions required to resolve GSI-191 and GL 2004-02 are addressed by the proposed TS change and supporting safety analysis discussed in Attachment 1 and summarized in the following paragraphs. Attachment 1 of this letter provides the North Anna plant-specific applications of the DOM-NAF-3 methodology for changes to the recirculation spray (RS) pump start method and the containment air partial pressure operating limits in TS Figure 3.6.4-1. The marked up and proposed TS pages are provided in Attachments 2 and 3, respectively. The associated marked up and typed Bases changes are provided for information only in Attachments 4 and 5, respectively.

The proposed TS change revises the method for starting the inside and outside RS pumps in response to a design basis accident. Currently the North Anna RS pumps start using delay timers that are initiated when the containment pressure reaches the Containment Depressurization Actuation (ODA) High High set point. The change will start the RS pumps with a coincident ODA High High pressure and refueling water storage tank (RWST) Level Low. Other changes required to support the containment Aco0'

Serial No.06-849 Docket Nos. 50-280/281 Proposed Technical Specification Change Page 2 of 5 reanalysis and ensure net positive suction head include: lowering the containment operating temperature limit, modifying the containment partial air pressure operating limit, and lowering the plant setpoint and TS allowable values for the RWST Level Low Low function that initiates safety injection recirculation mode transfer. In addition, the TS change modifies the sump inspection requirements to reflect the new strainer configuration.

The proposed safety analysis change revises the North Anna containment analyses by converting from the present Stone and Webster LOCTIC computer code to the GOTHIC code (Topical Report DOM-NAF-3). On August 30, 2006, the NRC staff approved DOM-NAF-3, which documents the Dominion methodology for analyzing the containment response to postulated pipe ruptures. Attachment 1 of this letter provides the North Anna plant-specific applications of the DOM-NAF-3 methodology for changes to the RS pump start method and the containment air partial pressure and temperature operating limits in TS Figure 3.6.4-1.

An additional safety analysis change includes revisions to the LOCA Alternate Source Term (AST) dose consequences analysis that accommodate the changes to the RS pump start methodology. The changes to the RS pump start methodology result in a short-term increase in air leakage from the containment and a short-term reduction in spray removal of radioactive isotopes from the containment atmosphere. Attachment 1 of this letter also documents revisions to the LOCA dose consequences analysis.

Dominion requests NRC staff approval of the proposed TS change and supporting safety analyses revisions by February 15, 2007 in order to implement the proposed changes during the spring 2007 refueling outage for North Anna Unit 2 and during the fall 2007 refueling outage for North Anna Unit 1. This schedule is necessary to meet the required implementation schedule for GSI-191/GL 2004-02 resolution. A staggered implementation of the TS change is required due to the plant modifications, which only can be performed during a plant outage. However, it is Dominion's intention to implement the North Anna Units 1 and 2 containment analyses with the GOTHIC code (replacing the Stone and Webster LOCTIC computer code) for both units during the spring North Anna Unit 2 refueling outage. Attachment 1 includes GOTHIC analyses for the current and proposed plant configurations. The current configuration analyses will be applicable to North Anna Units 1 and 2 upon NRC approval of the application of the GOTHIC methodology for North Anna.

Dominion has evaluated the proposed change and has determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. Dominion has also determined that operation with the proposed change will not result in a significant increase in the amount of effluents that may be released offsite or in a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion 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

Serial No.06-849 Docket Nos. 50-280/281 Proposed Technical Specification Change Page 3 of 5 change. The basis for these determinations is provided in Attachment 1.

The proposed changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee.

If you have any questions regarding this submittal, please contact Mr. Thomas Shaub at (804) 273-2763.

Very truly yours, Gearld T. Bischof Vice President - Nuclear Engineering Commitments made inthis letter: None Attachments: (5) - Discussion of Changes - Marked-up Technical Specification Pages - Proposed Technical Specifications Pages - Marked-up Technical Specification Bases Pages - Typed Technical Specifications Pages

Serial No.06-849 Docket Nos. 50-280/281 Proposed Technical Specification Change Page 4of 5 cc: U. S. Nuclear Regulatory Commission Regional Administrator - Region 11 Sam Nunn Atlanta Federal Center Suite 23 T85 61 Forsyth Street, SW Atlanta, Georgia 30303-893 1 Mr. S. R. Monarque Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 H 12 11555 Rockville Pike Rockville, MID 20852-2738 Mr. S. P. Lingam NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8 G9A Rockville, Maryland 20852-2738 Mr. J. T. Reece Senior Resident Inspector North Anna Power Station Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218

Serial No.06-849 Docket Nos. 50-280/281 Proposed Technical Specification Change Page 5 of 5 COMMONWEALTH OF VIRGINIA )

COUNTY OF HENRICO)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. He has affirmed before me that he 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 his knowledge and belief.

Acknowledged before me the 3L day ofJ 2 L 2006.

My Commission Expires:.A //,I '2iyo Notary Public

• -(SEAL)

Serial No.06-849 Docket Nos. 50-338/339 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 DISCUSSION OF CHANGES VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

Table of Contents Title Page Table of Contents........................................................................................ 2 List of Tables ............................................................................................ 4 List of Figures............................................................................................ 5 List of Acronyms and Abbreviations ................................................................... 6 1.0 Introduction........................................................................................ 7 2.0 Description of Changes ........................................................................... 8 2.1 Implement GOTHIC Containment Analysis Methodology......................................... 10 2.2 Start RS Pumps on RWST Level Low Coincident with High High Containment Pressure.......11 2.3 Change Containment Air Partial Pressure Operating Limits in TS Figure 3.6.4-1................. 18 2.4 Change Maximum Limit for Containment Temperature ........................................... 19 2.5 Change Automatic RMT Setpoint ................................................................... 19 2.6 LOCA Alternate Source Term ...... -... ***........ ................. 22 2.7 Containment Sump Surveillance Requirements..................................................... 23 3.0 GOTHIC Containment Analyses.............................................................. 24 3.1 Application of the GOTHIC Methodology........................................................ 26 3.1.1 Model Geometry................................................................................. 26 3.1.2 Engineered Safety Features ...................................................................... 27 3.1.3 Containment Passive Heat Sinks ................................................................ 28 3.1.4 Plant Parameter Design Inputs ................................................................. 29 3.1.5 Containment Initial Conditions and Instrument Uncertainty .................................. 29 3.1.6 NPSH Available and Water Holdup ........................................................... 30 3.2 Break Mass and Energy Release..................................................................... 35 3.2.1 LOCA Mass and Energy Releases............................................................... 35 3.2.2 MSLB Mass and Energy Releases .............................................................. 36 3.3 LOCA Peak Pressure and Temperature........................................................... 37 3.4 LOCA Containment Depressurization ............................................................ 40 3.5 LHSI Pump NPSH Analysis ....................................................................... 47 3.6 RS Pump NPSH Analysis .......................................................................... 55 3.7 MSLB Peak Pressure and Temperature........................................................... 68 3.7.1 MSLB Peak Pressure Analysis ................................................................ 68 3.7.2 MSLB3 Peak Temperature Analysis ............................................................ 69 3.8 Inadvertent QS Actuation Event................................................................... 74 3.9 EQ Envelope Verification .......................................................................... 75 3.10 Proposed TS Limits for Containment Air Partial Pressure vs. SW Temperature .......... 76 3.11 Summary of Containment Analysis Results..................................................... 78 4.0 Revised LOCA AST Analysis.................................................................. 83 4.1 Changes in Containment Pressure and Leakage Assumptions .................................. 84 4.2 Description of Containment Volumes.............................................................. 85 4.3 Changes in Containment Spray Removal Coefficients ........................................... 86 4.4 Changes in ECCS Leakage Assumptions ......................................................... 89 4.5 Changes in RWST Leakage Assumptions......................................................... 90 4.6 Changes in Control Room Occupancy Factors ................................................... 91 4.7 Timing of Release Phases ........................................................................... 91 Page 2

4.8 Control Room and Auxiliary Building Filter Efficiency.......................................... 92 4.9 Control Room Volume .............................................................................. 92 4.10 Revised Radiological Results ..................................................................... 92 5.0 Conclusions ...................................................................................... 93 6.0 References......................................................................................... 94 7.0 Regulatory Evaluation.......................................................................... 97 7.1 No Significant Hazards Consideration ............................................................ 97 7.2 Regulatory Requirements ......................................................................... 100 8.0 Environmental Assessment................................................................... 102 Page 3

List of Tables Table 3. 1-1: Key Parameters in the Containment Analysis .................................................. 32 Table 3.1-2: GOTHIC Model Heat Sink Material Properties .................................................. 34 Table 3.1-3: Containment Passive Heat Sinks ................................................................ 34 Table 3.3-1: LOCA Peak Pressure and Temperature Analysis Results ..................................... 38 Table 3.4-1: Containment Depressurization Results for Proposed Configuration............................... 44 Table 3.5-1: LHSI Pump NPSHa Analysis Results - Current Configuration................................. 49 Table 3.5-2: Time Sequence of Events for LHSI Pump NPSHa Analysis (Current Configuration) ......... 49 Table 3.5-3: LHSI Pump NPSHa Analysis Results - Proposed Configuration............................... 50 Table 3.54: Time Sequence of Events for LHSI Pump NPSHa Analyses - Proposed Configuration ......50 Table 3.6-1: Results for RS Pump NPSHa Analsyes (Current Configuration)............................... 58 Table 3.6-2: Time Sequence of Events for RS Pump NPSHa (Current Configuration) ...................... 58 Table 3.6-3: RS Pump NPSHa Results by Break and Single Failure for Proposed Configuration (10.3 psia, 35 F SW).......................................................................................... 59 Table 3.6-4: Time Sequence of Events from Limiting RS Pump NPSHa. Analyses (Proposed Configuration)...................................................................................... 60 Table 3.7- 1: Results from MSLB Containment Peak Pressure Analyses...................................... 70 Table 3.7-2: Time Sequence of Events from MSLB Peak Pressure Analysis- Proposed Configuration.....70 Table 3.7-3: Results from MSLB Containment Peak Temperature Analyses.................................. 70 Table 3.11 -1: GOTHIC Containment Analysis Results...................................................... 79 Table 3.11-2: Matrix of Conservative Inputs for North Anna GOTHIC Containment Anal yses............ 80 Table 4. 1-1: Containment Leak Rate Assumption............................................................ 85 Table 4.2-1: Time Dependent Sprayed/Unsprayed Containment Fractions.................................. 85 Table 4.3-1: Spray System Characteristics................................................................... 86 Table 4.3-2: Current Combined QS and RS Aerosol Removal Coefficients................................. 87 Table 4.3-3: Aerosol Removal Coefficients .................................................................. 89 Table 4.4-1: Containment Sump Volume vs. Time ........................................................... 90 Table 4.10-1: Revised Design Basis LOCA Dose Results................................................... 92 Page 4

List of Figures Figure 2. 1-1: North Anna RWST Level Low ESFAS Initiation ...................................................... 17 Figure 2.1-2: North Anna RWST Level Low-Low ESFAS Initiation ................................................ 21 Figure 3.3-1: Comparison of Containment Pressure from DEHLG Peak Pressure Analysis......................... 39 Figure 3.3-2: Containment Vapor Temperature from DEHLG Peak Pressure Analysis ............................. 39 Figure 3.4-1: Comparison of Containment Pressure from DEPSG Depressurization Analysis...................... 45 Figure 3.4-2: Comparison of Containment Temperature from DEPSG Depressurization Analysis ................. 45 Figure 3.4-3: Comparison of Total RSHX Heat Rate from DEPSG Depressurization Analysis .................... 46 Figure 3.5-1: LHSI Pump NPSHa -Cu'rrent Configuration (9.0 psia, 95 F).......................................... 51 Figure 3.5-2: Containment Pressure from LHSI Pump NPSHa Analysis - Current Configuration .................. 51 Figure 3.5-3: Containment Temperature from LUSI Pump NPSHa Analysis - Current Configuration ............. 52 Figure 3.5-4: Total RSHX Heat Rate from LHSI Pump NPSHa Analysis - Current Configuration ................ 52 Figure 3.5-5: LHSI Pump NPSHa - Proposed Configuration (10.3 psia, 75 F) ..................................... 53 Figure 3.5-6: Containment Pressure from LHSI Pump NPSHa Analysis - Proposed Configuration ................ 53 Figure 3.5-7: Containment Temperature from LHSI Pump NPSHa Analysis - Proposed Configuration............ 54 Figure 3.5-8: Total RSHX Heat Rate from LUSI Pump NPSHa Analysis - Proposed Configuration............... 54 Figure 3.6-1: IRS Pump NPSHa - Current Configuration (8.85 psia, 73 F) ......................................... 61 Figure 3.6-2: Containment Pressure from IRS Pump NPSHa Analysis - Current Configuration.................... 61 Figure 3.6-3: Containment Temperature from IRS Pump NPSHa Analysis- Current Configuration................ 62 Figure 3.6-4: Total RSHX Heat Rate from IRS Pump NPSHa Analysis- Current Configuration................... 62 Figure 3.6-5: ORS Pump NPSHa - Current Configuration (8.85 psia, 73 F) ........................................ 63 Figure 3.6-6: IRS Pump NPSHa - Proposed Configuration (10.3 psia, 35 F)........................................ 64 Figure 3.6-7: Containment Pressure from IRS Pump NPSHa Analysis - Proposed Configuration .................. 64 Figure 3.6-8: Containment Temperature from IRS Pump NPSHa Analysis - Proposed Configuration ............. 65 Figure 3.6-9: Total RSHX Heat Rate from IRS Pump NPSHa Analysis - Proposed Configuration ................ 65 Figure 3.6-10: ORS Pump NPSHa - Proposed Configuration........................................................ 66 Figure 3.6-11: Containment Pressure from ORS Pump NPSHa Analysis - Proposed Configuration ............... 66 Figure 3.6-12: Containment Temperature for ORS Pump NPSHa Analysis - Proposed Configuration ............. 67 Figure 3.6-13: Total RSHX Heat Rate from ORS Pump NPSHa Analysis - Proposed Configuration.............. 67 Figure 3.7-1: Containment Pressure from 30% Power, 1.4 ft2 MSLB Peak Pressure Analysis - Proposed Configuration........................................................................................................... 71 Figure 3.7-2: Containment Temperature from 102% Power, 0.6 ft2 MSLB Peak Temperature Analysis -

Proposed Configuration ................................................................................................ 71 Figure 3.7-3: Containment Pressure Comparison from MSLB Peak Temperature Analyses -

Proposed Configuration ............................................................................................... 72 Figure 3.7-4: Comparison of MSLB Peak Temperature Analyses - Proposed Configuration........................ 73 Figure 3. 10- 1: Containment Air Partial Pressure versus Service Water Temperature (Proposed TS Figure 3.6.4-1)......................................................................................... 77 Page 5

List of Acronyms and Abbreviations ADF Atmospheric Dispersion Factor AFW Auxiliary Feedwater AST Alternate Source Term CDA Consequence Depressurization Actuation CDT Containment Depressurization Time COT Channel Operational Test DEHLG Double Ended Hot Leg Guillotine DEPSG Double Ended Pump Suction Guillotine DER Double Ended Rupture DLM Diffusion Layer Model DPP Depressurization Peak Pressure EAB Exclusion Area Boundary ECCS Emergency Core Cooling System EQ Equipment Qualification ESF Engineered Safety Features ESFAS Engineered Safety Features Actuation System HiHSI High Head Safety Injection IRS Inside Recirculation Spray LHSI Low Head Safety Injection LOCA Loss of Coolant Accident MSLB Main Steam Line Break NAPS North Anna Power Station NPSHa Available Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supplý System ORS Outside Recirculation Spray PCT Peak Clad Temperature PWR Pressurized Water Reactor QS Quench Spray RCS Reactor Coolant System RMT Recirculation Mode Transfer RS Recirculation Spray RSHX Recirculation Spray Heat Exchanger RWST Refueling Water Storage Tank SBLOCA Small Break LOCA SG Steam Generator SI Safety Injection SRP Standard Review Plan SW Service Water TS Technical Specifications UFSAR Updated Final Safety Analysis Report Page 6

1.0 Introduction Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests an amendment to Facility Operating License Numbers NPF-4 and NPF-7 in the form of changes to the Technical Specifications (TS) for North Anna Power Station Units 1 and 2 and the current approved containment analysis methodology.

This report documents the implementation of changes to the North Anna Power Station (NAPS) plant safety analyses to support the resolution of NRC Generic Letter 2004-02 [1]. Section 2 describes the changes to the plant licensing bases that are necessary to support the containment sump strainer replacement project. Section 3 summarizes the GOTHIC containment analyses using the methodology in topical report DOM-NAF-3 [3, 27]. GOTHIC analyses were performed for both the current plant configuration and a "proposed configuration" that includes changes described in Sections 2.2 (RS pump start using RWST level), 2.3 (increase to containment air partial pressure limits), 2.4 (reduction in maximum containment temperature limit), and 2.5 (change in setpoint for automatic recirculation mode transfer (RMT) for the safety injection system). The GOTHIC analyses represent a change to a UFSAR method of evaluation, as defined in 10CFR5O.59, for NAPS. The NRC approved DOM-NAF-3 in Reference 27. Section 4 documents revisions to the LOCA Alternate Source Term (AST) analysis that are required to support the delayed start of the RS pumps and the increase to the containment air partial pressure limits. Section 2.6 describes the licensing basis changes. Section 5 documents the conclusions of the GOTHIC and LOCA AST analyses. References are listed in Section 6. Section 7 documents the Regulatory Evaluation and Section 8 documents the Environmental Assessment.

The proposed changes qualify for categorical exclusion for an environmental assessment as set forth in 10 CER 51.22(c)(9). Therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

Page 7

2.0 Description of Changes Changes to the North Anna Power Station (NAPS) licensing bases and Technical Specifications (TS) are proposed to support resolution of NRC Generic Letter 2004-02 [1]. In a letter dated September 1, 2005, Dominion (Virginia Electric and Power Company) included three commitments for NAPS (numbers 6-8 in Attachment 6) to resolve NRC GL 2004-02 [2]. The commitments are repeated with a brief discussion about how each is resolved in this report.

o Dominion will report tire minimum NPSH margin in the NAPS-plant-specific LAR described in Item 2(e).

Table 3.11-1 summarizes the minimum NPSH available for the low head safety injection (LHSI), inside recirculation spray (IRS), -and outside recirculation spray (ORS) pumps using the GOTHIC containment analysis methodology described in DOM-NAF-3. NPSH margins are reported for two sets of analyses that are described in Section 3. The "proposed configuration" NPSH margins (available NPSH - required NPSHI) are the design values for the containment sump strainer project.

u Dominion will submit the GOTHIC containment analysis methodology w~ith plant-specific analyses that support the proposed changes to TS Figure 3.6 4-1 and the RS pump start method in February2006.

The GOTHIC containment analysis methodology was submitted to the NRC as topical report DOM-NAF-3 [3] in a letter dated November 1, 2005 [4]. A supplement to DOM-NAF-3 was submitted in a letter dated July 14, 2006 [17]. The NRC issued the Safety Evaluation Report for DOM-NAF-3 on August 30, 2006 [27]. The containment analyses in Section 3 apply the DOM-NAF-3 methodology without modification to demonstrate compliance with containment design criteria. The proposed change to start the RS Pumps using RWST level is described in Section 2.2. The proposed change to TS Figure 3.6.4-1 is described in Section 2.3.

o The planned changes to delay the RS pumps and to modify TS Figure 3.64-1 require a relaxation of the currently approved containment leakage assumptionsfor NAPS. Dominion will submit a revisedASTLOCA analysisfo~r NA PSfor NR C review in February2006.

Section 3 describes how the GOTHIC LOCA containment analyses for the proposed plant configuration are subatmospheric within one hour but the containment pressure exceeds 0.5 psig during hours 1-4 but is less than 2.0 psig during hours 1-6 and is subatmospheric within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The current NRC-approved containment leakage assumption in the LOCA Alternate Source Termn (AST) analysis corresponds to a containment pressure of 0.5 psig during hours I-

4. Section 2.6 describes the changes to increase the allowable containment leakage to 2.0 psig during the time interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the LOCA initiation.

Page 8

Since our September 1, 2005 letter, Dominion has identified additional license amendments that are required to resolve NRC GL 2004-02. These license amendments are detailed later in Section 2:

u Change TS maximum limit for containment temperature; o Change TS allowable values for safety injection (SI) automatic recirculation mode transfer (RMT); and o Change TS surveillance requirements for the containment sump to be consistent with the planned design for separate strainers for the SI and RS systems.

o Change TS 5.5.15 value for Pa, the peak calculated containment pressure from a LOCA, based on the GOTHIC containment analyses in Section 3.3.

The affected TS Bases will be changed to be consistent with the TS changes and the revised safety analyses. TS Bases changes, reflecting the proposed change with the Technical Specification change discussed above, are included for information only. The TS Bases will be revised in accordance with the TS Bases Control Program, TS 5.5.13 following NRC approval of the license amendment.

Page 9

2.1 Implement GOTHIC Containment Analysis Methodology The current licensing basis analysis methodology for loss of coolant 'accident (LOCA) containment response is the Stone & Webster LOCTIC computer code that is described in NAPS UFSAR Chapter 6 [5]. The LOCTIC methodology will be replaced with the GOTHIC analytical methodology that is described in topical report DOM-NAF-3 [3]. The topical report was submitted to the NRC for review and approval on November 1, 2005 [4], with a supplement submitted to the NRC on July 14, 2006 [17], in advance of this license amendment request. The NRC Safety Evaluation Report for DOM-NAF-3 was issued on August 30, 2006 [27]. The GOTHIC design analyses summarized in this report have used the topical report methodology without modification and consistent with Reference 27. The GOTIHIC analyses in Section 3 replace the LOCTIC analyses in NAPS UFSAR Chapter 6 for calculation of the following containment design requirements:

1. LOCA peak containment pressure and temperature,

2. LOCA containment depressurization time,

3. LOCA containment peak pressure following depressurization,

4. Available net positive suction head (NPSHa) for the LHSI pumps,

5. NPSHa for the ORS and IRS pumps, and

6. Main steam line break (MSLB) peak containment pressure and temperature GOTHIC is used to verify that the containment liner temperature is less than the limit using the methodology in Section 3.3.3 of DOM-NAF-3 for LOCA and MSLB events. GOTHIC also can be used to verify equipment temperatures within design limits using the methodology in Section 3.3.4 of DOM-NAF-3. Finally, the minimum containment water level and maximum sump liquid temperatures from GOTHIC NPSH calculations will be used to establish bounding inputs to the sump strainer design. The GOTHIC NPSH analysis methodology in Section 3.8 of DOM-NAF-3 ensures a conservative prediction of minimum containment water level (i.e., accounting for water holdup) and maximum sump liquid temperature. Depressurization analyses are biased to maximize the total pressure and provide a conservative minimum sump water temperature. In conclusion, the GOTIHIC methodology for long-term analysis of NPSH and containment depressurization ensures conservative results for component design (e.g., strainer debris head loss and component stress analyses).

Page 10

2.2 Start RS -Pumpson RWST Level Low Coincident with High High Containment Pressure NAPS is a three-loop Westinghouse PWR with a subatmospheric containment design. The following plant description is consistent with Chapter 6 of the NAPS UFSAR. The engineered safety features (ESF) that mitigate a LOCA or MSLB event include:

" A safety injection (SI) system that injects borated water into the cold legs of all three reactor coolant loops for the entire spectrum of reactor coolant system (RCS) break sizes to limit core temperature, maintain core integrity, and provide negative reactivity for additional shutdown capability.

o Two separate low-head safety injection (LHSI) subsystems, either of which provides long-term removal of decay heat from the reactor core.

" Two separate subsystems of the containment depressurization system--quench spray (QS) and recirculation spray (RS)-that operate together to reduce the containment temperature, return the containmen *tpressure to subatmospheric, and remove heat from the containment. The RS subsystem maintains the containment subatmospheric and transfers heat from the containment to the service water (SW) system.

The QS system consists of two pumps that start on a Containment Depressurization Actuation (CDA)

High High containment pressure signal and draw suction from the refueling water storage tank (RWST) until the tank is empty. The RS system consists of four independent trains, each with one pump that takes suction from the containment sump. Two inside recirculation spray (IRS) pumps are located inside the containment sump, while two outside recirculation spray (ORS) pumps are located in the Safeguards Building. Currently, the RS pumps are started using delay timers that are initiated on the CDA signal. Each RS train has a recirculation spray heat exchanger (RSHX) that is cooled by SW (on the tube side) for long-term containment heat removal. The casing cooling subsystem includes two pumps that start on a CDA signal and take suction from the casing cooling tank. Each casing cooling pump discharges cold water to the suction of its respective ORS pump to increase the available NPSH.

The SI system consists of two LHSI pumps and three HHSI pumps that draw from the RWST and inject into the RCS cold legs. The SI pumps take suction from the RWST until a low-low level is reached, at which time recirculation mode transfer (RMT) occurs. The RMT function changes the LHSI pump suction from the RWST to the containment sump and the HHSI pump suc tion from the RWST to the discharge header of the LHSI pumps.

Because the RS and SI systems use the containment sump to demonstrate that design criteria are satisfied, the resolution of NRC Generic Letter 2004-02 affects the IRS, ORS and LUSI pumps.

Section 3.7.2.3.2.4 of NEI-04-07 [6] has different requirements for demonstrating adequate pump performance whether the sump strainer is fully or partially submerged when the LHSI and RS pumps are operating. For a fully submerged strainer, the strainer debris head loss must be less than or equal to Page I11

the NPSH margin. For a partially submerged strainer, the strainer debris head loss must be less than one-half the pool height. Thus, if the strainer submergence height is only 1.0 ft at pump start, then the allowable debris head loss is 0.5 ft. further, only the wetted strainer surface area can be credited, and these limitations together could impose a very large strainer footprint.

Curren tly, the NAP S RS pumps start using delay timers that are initiated when the containment pressure reaches the CDA High Hi-gh containment pressure setpoint. The IRS pumps have a 400-second setpoint and the ORS pumps have a 210-second setpoint. At these start times, the containment water level is predicted to be less than 1 ft in the current UFSAR containment analyses. While there is sufficient NPSH margin for the pumps, the current timer delay setpoints start the RS pumps when the sump strainer is partially submerged. Because the partial submergence requirement may be too restrictive for the sump strainer design, the RS pump start will be delayed until sufficient water level is available in the containment.

Proposed Modification NAPS proposes to start the IRS and ORS pumps on 60% RWST wide range (WR) level coincident with a CDA H~igh High containment pressure signal. The ORS pumps will receive an immediate start signal once the coincidence logic is satisfied. The IRS pumps will start using a 120-second delay timer from the coincident actuation signal. This delay will minimize the impact on emergency diesel generator loading and allow for the ORS system to fill its piping completely, deliver spray to the containment, and reach a stable flow demand on the sump before the IRS pumps start. This method of starting the RS pumps ensures that a reliable mass of liquid has been added to the containment to meet the sump strainer submergence' requirements for the range of LOCA break sizes' that require the containment sump. The use of RWST WVR level to start the RS pumps classifies the new instrumentation as part of the Engineered Safety Features Actuation System (ESFAS). Thus, the design will include safety-grade instrumentation consistent with UFSAR Section 7.3, "Engineered Safety Features Actuation System"i7 with allowable values and surveillances that must be added to the NAPS Technical Specifications [7].

Plant Safety Analysis Impact Delaying the start method of the RS pumps has a potential adverse impact on several design criteria, including LOCA and MSLB containment pressure and temperature, environmental conditions for safety-related equipment inside containment, diesel loading, and dose consequences analyses. The following impacts on the NAPS UPSAR safety analyses and design were evaluated.

c3 Less energy is removed from the containme-nt sump liquid before the LHSI pumps take suction from the containment sump, causing a decrease in NPSHa for the LHSI pumps. The LHSI pump NPSHa analysis determnines the lowest acceptable RWST water level setpoint for RS pump initiation. Section 3.6 demonstrates that the LHSI pumps have sufficient NPSH margin with the proposed change to start the IRS and ORS pumps on 60% RWST WIR level, provided that other Page 12

changes are made to offset the LUSI pump NPSH margin reduction from delaying the RS pump start. Specifically, the safety analyses require an increase in containment air pressure, a change to the RMT actuation setpoint, and a reduced containment temperature operating limit of 115 F.

o The LOCA and MSLB containment pressures and temperatures are higher during the period when only the QS system is delivering spray flow to containment. The GOTHIC analyses in Sections 3.3 and 3.4 show the LOCA containment pressure and temperature decreasing before the ORS pumps start at approximately 30 minutes (assuming 1 train of ESF), and the RS pump start increases the depressurization rate. The delayed RS pump start increases the containment depressurization time to subatmospheric conditions. Section 3.4 demonstrates that the containment pressure after a LOCA is less than 2.0 psig within one hour and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The current LOCA dose consequences analysis assumes a containment leak rate at the TS 5.5.15 limit of 0. 1%

of containment volume per day for the first hour of the accident, a conservative leak rate corresponding to 0.5 psig containment pressure for the time interval from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and no leakage after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (i.e., containment pressure is subatmospheric) [20, Section 3.1.1.3]. The proposed changes create containment pressure responses that are not bounded by the current AST analysis assumption for containment leakage. Thus, Section 2.6 describes LOCA AST analyses that increase the containment leak rate during the time interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to a value that corresponds to a containment pressure of 2.0 psig. The GOTHIC analyses in Section 3.7 show the MSLB containment pressure and temperature without any credit for the RS systemr. The QS pump is the lone spray pump running against a SG boiloff rate of the maximum AFW flow rate.

Containment pressure and temperature increase to a peak at 30 minutes when the AFW flow to the faulted SG is terminated, after which the pressure aind temperature decrease based on the cooling capacity of the QS pump.

o The delayed RS pump start creates more adverse pressure and temperature conditions for the operation of safety-related equipment inside containment. The GOTHIC analyses in Section 3 created LOCA and MSLB containment pressure and temperature profiles based on the proposed change to the RS pump start. Various break locations, break sizes, and single-failures were considered. Section 3.9 concludes that the GOTHIC profiles remain bounded by existing equipment qualification test temperature and pressure data or are acceptable by evaluation.

Therefore, the operation of safety-related equipment inside containment during a LOCA or MSLB is not affected adversely by the proposed change to delay starting the RS pumps.

o Starting the RS pumps later can reduce the removal of iodine from the containment atmosphere and potentially increase the release of iodine to the environment during the period when RS is not operating. The current LOCA AST basis starts to credit iodine removal when outside RS pumps start at 289 seconds. Section 2.6 documents a change to the LOCA AST bases to credit iodine removal from the RS system at 40 minutes post-LOCA.

Page 13

o3 Currently, the RS pumps start using time delays from the CDA actuation signal (210 seconds for ORS and 400 seconds for IRS). Using 60% RWST WR level coincident with High High containment pressure delays the RS pump start until at least 14 minutes after accident initiation.

This pump start delay reduces the early loads on the emergency diesel generator. The ORS pumps will receive an immediate start signal once the coincidence logic is satisfied. The IRS pumps will start using a 120-second delay timer from the coincident actuation signal. This delay is sufficient to avoid simultaneous starting of the RS pumps on the same emergency diesel generator. Thus, the proposed change does not have an adverse impact on emergency diesel generator capability.

o3 The NPSHa for the RS pumps increases. Section 3.6 demonstrates how the ORS and IRS pumps have more NPSH margin from higher containment water levels and subcooling of the sump water by casing cooling and bleed flow from the QS system that injects at the IRS pump suctions.

Sections 3 and 4 of this report demonstrate that the NAPS safety analyses satisfy the accident analysis acceptance criteria and other design requirements when the RS pumps are started on 60% RWST WR level coincident with a High High containment pressure, provided that other changes are made to the containment air partial pressure operating limits (Section 3.10), maximum containment te mperature limit, and SI automatic recirculation mode transfer setpoint to provide sufficient NPSH margin.

Changes to NAPS Technical Specifications The use of RWST WR level to start the RS pumps classifies the new instrumentation as part of the ESFAS. The allowable values and surveillances for the ESFAS function must be added to the NAPS Technical Specifications [7]. The signals from the RWST WR level channels are used to initiate RMT for the safety injection system. Three of the level channels will be used for the RS pump start circuitry.

RMT occurs when RWST WR level reaches a "Low Low" setpoint according to TS 3.3.2. Since the RWST level setpoint for initiation of the RS pumps will be a higher value (60%) than the proposed RMT setpoint (16.0%), the RS pump initiation function will use the term "RWST Level Low".

The RWST Level Low trip will be designed to de-energize to actuate. The RS pump start will depend on a coincidence of 60% RWST WvR level with Containment Pressure HI-gh High (which is 2-out-of-4 channels energize to actuate), so a spurious de-energization of the RWST level circuits would not cause an unnecessary start of the RS pumps. TS 3.6.7 requires the RS system to be operable in Modes 1 through 4, but automatic system actuation is only required to be operable in Modes 1, 2 and 3 per TS 3.3.2. The RWST Level Low channels shall be operable in Modes 1 through 3 to be consistent with the Containment Pressure High-High channels (TS 3.3.2 Function 2.c). The RWST Level Low trip will actuate from 2-out-of-3 channels and I1-out-of-2 trains. The 2-out-of-3 logic configuration is consistent with other ESFAS circuits that de-energize to actuate at North Anna. The proposed change to TS 3.3.2 requires an inoperable channel to be placed in trip, leaving a I -out-of-2 configuration that satisfies redundancy requirements. Condition D in TS 3.3.2 specifies the required actions for an inoperable Page 14

channel for an ESFAS function that is required in MODES 1, 2, and 3 and is selected for the RWST Level Low channels. The surveillance requirements identified in the TS changes are consistent with the requiremvents for the SI RMT instrumentation (ESFAS Function 7.b). The site-specific PRA analysis was reviewed to confirm that the signal unavailability and risk impacts from the proposed RWST Level Low channels surveillance frequencies and completion times are consistent with those evaluated in WCAP-14333-P-A [26] and approved by the NRC.

Manual initiation of recirculation spray is required in Mode 4, even though automatic actuation is not required. In Mode 4, adequate time is available to manually actuate required components in the event of a DBA. However, because of the large number of components actuated on a CDA, actuation is simplified by the use of the manual actuation switches (ESFAS Function 2.a in TS 3.3.2). To be consistent with ESFAS function 2.b for containment spray, the automatic actuation logic and actuation

• relays for RWST Level Low shall be operable in Mode 4.

The determination of TS Allowable Values for the NAPS RWST Level Low initiation used Method 1 in ISA-RP67.04.02-2000 [20]. There are two Analytical Limits and thus two Allowable Values associated with this new function. The Analytical Limits are > 57.50 % WR Level and < 62.50 % WR Level. The corresponding Allowable Values are > 59.00 % WR Level and < 61.00 % WR Level from the following analysis. The reader is referred to Figure 2.1-1 for the relationship between the Analytical Limits and the Allowable Values.

Adding the Total Loop Uncertainty to the Analytical Limit yields a Minimum Trip Setpoint of 59.24 % WR Level. Adding the Non-COT (Channel Operational Test) error components t o the Analytical Limit yields a Minimum Allowable Value of 58.868 % WR Level. The Actual Nominal Trip Setpoint of 60.00 % WR Level is conservative with respect to the Minimum Trip Setpoint.

The Actual Allowable Value of > 59.00 % WR Level is conservative with respect to the Minimum Allowable Value. The Allowable Value of > 59.00 % WYR Level is based on maintaining a Nominal Trip Setpoint value of 60.00 % WR Level. The proposed Allowable Value of > 59.00 %

WR Level is conservative with respect to the calculated value using rack error terms (i.e., COT error terms).

Subtracting the Total Loop Uncertainty from the Analytical Limit yields a Maximum Trip Setpoint of 60.8 85 % WR Level. Subtracting the Non-COT error components from the Analytical Limit yields a Maximum Allowable Value of 61.26 % WR Level. The Actual Nominal Trip Setpoint of 60.00 % WR Level is conservative with respect to the Maximum Trip Setpoint. The Actual Allowable Value of < 61.00% WvR Level is. conservative with respect to the Maximum Allowable Value. This Allowable Value of < 61.00 % WR Level is based on maintaining a Nominal Trip Setpoint value of 60.00 % WR Level. The proposed Allowable Value of < 61.00 %

WR Level is conservative with respect to the calculated value using rack error terms (i.e., COT error terms).

Page 15

The statistical combination of the COT and Non-COT error components is provided below and used in Figure 2.1-1 to determine the Minimum/Maximum Trip Setpoints and the Minimum/YMaximum Allowable Value s.

Non-COTe,,or = SE + [EA 2 + PMA2 + PEA' + (SCA+SMTE) 2 + SD 2 + SPE 2 + STE' + SPSE 2 +

2 M1MTE 2 + M3MTE +RT2]1 Non-COT,,or = 0.064 + [0.02 + 0.022 2 +0.02 + (0.5+0.2 11)2 + 0.222 2 +0.02 + 0.9332 + 0.02 +

0.153 2 + 0.03 2 + 0.5 2] 1/2 Non-COTei.ro = + 1.368 % of span and - 1.24 % of span 2

COTeot= +/- (M1 2 +M3 2 + RD ) 1/2 COTerror = +/- (0.12 +0.25 2+ 1.02)12 COTermro=+/-+1.036 % of span Based on the above evaluations, the following changes to the NAPS Technical Specifications are proposed for implementation of the RWST Level Low instrumentation to start the IRS and ORS pumps on 60% RWST WR level coincident with High High containment pressure:

o TS Table 3.3-2: Add the RWST Level Low Coincident with High High Containment Pressure as ESFAS Function 2.d with the requirement for 2-out-of-3 channels to trip and 1-out-of-2 ESFAS trains to trip. The ESFAS logic is required in Modes 1, 2 and 3 consistent with the requirements for Containment Pressure High-High (ESFAS Function 2.c). Condition D is selected for the RWST Level Low channels and Condition C applies to the Automatic Actuation Logic and Actuation Relays (ESFAS Function 2.b). RWST Level Low coincident with High High Containment Pressure has Allowable Values of > 59% and < 61% WR level.

Page 16

Figure 2.1-1: North Anna RWST Level Low ESFAS Initiation NORTH ANNA'S RWST LEVEL LO ESFAS INITIATION Nominal Operating Limit 97.6 % WR Level Low Operating Limit 96.7 % WR Level High Analytical Limit (AL) 62.50 % WR Level 0 00 Maximum Allowabie Value (MAy)

OPERATING oC r 61.26% WR Level MARGIN I- - -

Lo 0 jo Actual Allowable Value (AV) 36.7 % WR Level z 61.0 % WR Level Maximum Trip Setpoint (MTS) 60.89 % WR Level SAFETY MARGIN 0.89% WR Level 60 Actual Trip Setpoint (ATS) 60.0 %WR Level SAFETY MARGIN 0.76 % WR Level 00 Minimum Trip Setpolnt (MTS) 59.24 % WR Level

5. CL 0 Actual Allowable Value

-JR 59.0 % WR Level T- Minimum Allowable Value 58.87 %WR Level z

Z UJ Low Analytical Lim It (AL) 57.50 % WR Level Note: The COT errors are based on the Minimum Trip Setpoint value minus the Minimum Allowable value and the Actual Trip Setpoint value minus the Actual Allowable Value.

Page 17

2.3 Change Containment Air Partial Pressure Operating Limits in TS Figure 3.6.4-1 The GOTHIC containment analyses for LOCA containment peak pressure (Section 3.2), MSLB containment peak pressure (Section 3.7), and LOCA containment depressurization (Section 3.4) support an increase in the containment air partial pressure upper limit in TS Figure 3.6.4-1. The proposed TS upper limit is 12.3 psia from 35 F to 55 F SW temperature, linearly decreasing to 10.4 psia at 95 F. The TS lower limit will be increased to a constant 10.3 psia to recover LHSI pump NPSH margin that has been reduced by the delayed RS pump start. The proposed change to TS Figure 3.6.4-1 and the technical basis for the upper and lower limits are presented in Section 3. 10 of this report.

The combined increase to the air partial pressure limits and the delayed start of the RS pumps create analyzed containment pressures that are greater than 0.0 psig after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for double-ended pump suction guillotine (DEPSG) breaks with one train of emergency safety features. Section 3.4 shows the DEPSG break proposed configuration analyses produce a containment pressure that is greater than 0.0 psig but less than 2.0 psig during 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 /> after the event initiation. The GOTHI1C analyses predict containment pressures that are less than 0.0 psig after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To accommodate the increased containment pressure profiles, the containment leakage in the LOCA dose consequences analysis was increased to correspond to 2.0 psig during 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 /> (see Section 2.6).

As a result, the affected TS Bases are changed from the current analyzed pressure of 0.5 psig from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 2.0 psig from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the LOCA.

Page 18

2.4 Change Maximum Limit for Containment Temperature Currently, the containment average temperature is limited to 120 F by TS 3.6.5 and TS Figure 3.6.4-1.

The maximum containment temperature is a limiting initial condition for several of the LOCA and MSLB analyses, as shown in Table 3.11-2. North Anna proposes to reduce the TS maximum temperature limit from 120 F to 115 F to recover design margin. With a lower initial containment temperature, the passive heat sinks can condense more steam in the early phases of a LOCA or MSLB event, generating lower pressures in containment. For long-term analyses, a lower initial containment temperature reduces the amount of initial stored energy in the containment passive heat sinks that is eventually passed to the containment atmosphere, quenched by spray flow, and added to the containment sump. Thus, the long-term NPSH analyses see a small benefit. The proposed reduction in the TS containment temperature limit does not create a burden on the plant, as containment average temperatures typically are less than -105 F.

2.5 Change Automatic RMT Setpoint North Anna proposes to change the safety injection automatic recirculation mode transfer (RMT) setpoint from 19.4 % to 16.0 % RWST wide range level. The purpose of the change is to delay the time of RMT so that the RS system can remove more energy from the containment and reduce the sump temperature before the HSIS pumps swap suction from the RWST to the containment sump.

Delaying RM~T also provides a higher containment water level before LHSI swapover to recirculation.

The lower sump temperature and higher water level increase the NPSH margin for the LHSI pumps at RMIT. The containment analyses in Section 3 reflect the change and include a 2.5 % uncertainty on the RMT setpoint. The current configuration analyses assume a range of 16.9-21.9 % WR level, while the proposed configuration analyses assume a range of 13.5-18.5 %WR level. For the LHSI pump NPSH analyses in Section 3.6, the setpoint change increases the RMT initiation time by 212 seconds for the cases analyzed at 95 F SW. The proposed setpoint of 16.0 % WR level provides sufficient time to complete the automatic valve manipulations with margin to preclude air entrainment in the SI piping System.

Changes to NAPS Technical Specifications The SI RMT function occurs when RWST level reaches the "RWST Level Low-Low" setpoint. The current plant setpoint is 19.4 % WR level and the TS Allowable Values are > 18.4 % and < 20.4 % in TS Table 3.3.2-1. The determination of TS Allowable Values for the NAPS RWST Level Low-Low initiation used Method 1 in ISA-RP67.04.02-2000 [20]. There are two Analytical Limits and thus two Allowable Values associated with this new function. The Analytical Limits are > 13.50

% WR Level and < 18.50 % WR Level. The corresponding Allowable Values are > 15.00 % WR Level and < 17.00 % WR Level from the following analysis. The reader is referred to Figure 2.1-2 for the relationship between the Analytical Limits and the Allowable Values.

Page 19

Adding the Total Loop Uncertainty to the Analytical Limit yields a Minimum Trip Setpoint of 15.243 % WR Level. Adding the Non-COT (Channel Operational Test) error components to the Analytical Limit yields a Minimum Allowable Value of 14.868 % WR Level. The Actual Nominal Trip Setpoint of 16.00 % WR Level is conservative with respect to the Minimum Trip Setpoint.

The Actual Allowable Value of > 15.00 % WR Level is conservative with respect to the Minimum Allowable Value. The Allowable Value of > 15.00 % WR Level is based on maintaining a Nominal Trip Setpoint value of 16.00 % WR Level. The proposed Allowable Value of > 15.00 %

WR Level is conservative with respect to the calculated value using rack error terms (i.e., COT error terms).

Subtracting the Total Loop Uncertainty from the Analytical Limit yields a Maximum Trip Setpoint of 16.885 %WR Level. Subtracting the Non-COT error components from the Analytical Limit yields a Maximum Allowable Value of 17.26 % WR Level. The Actual Nominal Trip Setpoint of 16.00 % WR Level is conservative with respect to the Maximum Trip Setpoint. The Actual Allowable Value of < 17.00%. WR Level is conservative with respect to the Maximum Allowable Value. This Allowable Value of < 17.00 % WR Level is based on maintaining a Nominal Trip Setpoint value of 16.00 % WR Level. The proposed Allowable Value of < 17.00 %

WR Level is conservative with respect to the calculated value using rack error terms (i.e., COT error terms).

.The statistical combination of the COT and Non-COT error components is provided below and used in Figure 2.1-2 to determine the Minimum/Maximum Trip Setpoints and the Minimum/Maximum Allowable Values.

Non-COTemor = SE + [EA 2 + PMA 2 + PEA 2 + (SCA+SMTE) 2 + SD2 + SpE 2 + STE 2 + SpSE 2 +

2 2 M1MTE 2 + M3MTE 2 + RTE ] 1 Non-COTennr =0.064 + [0.02 + 0.022 2 +0.02 + (0.5+0.2 11)2 + 0.2222 + 0.0 2 + 0.933 2 +0.02 +

0.153 2 + 0.03 2+ 0.5 2] 1/2 Non-COTcror = + 1.368 % of span and - 1.24 % of span COTe.r,,, = + (Mi1' +M3 2 + RD2) 112 COTerror = +/- (0.12 +0.25 2 + 1.02) 1/2 COTe...or = +/- 1.036 % of span Based on the above evaluations and the safety analyses in Section 3, it is proposed to change the NAPS Technical Specifications RWST Level Low-Low function Allowable Values to > 15% and < 17%

RWST WR level.

Page 20

Figure 2.1-2: North Anna RWST Level Low-Low ESFAS Initiation NORTH ANNA'S RWST LEVEL LO-2 ESFAS INITIATION Nominal Operating Limit 97.6 % WR Level Low Operating Limit 96.7 % WR Levei High Analytical Limit (AL) 18.50 % WR Level 0b 90.

z ~Z L Maximum Allowable Value (MAy)

OPERATING 17.26 % WR Level MARGIN -0 cc Actuai Allowable Value (AV) 80.7 % WR Level 17.0 % WR Level Maximum Trip Setpoint (MTS) 16.89 % WR Level 00 SAFETY MARGIN Actual Trip SetpoInt (ATS) 0.76 % WR Level 16.0 % WR Level 0o CC Minimum Trip Setpoint (MTS)

TI 15.24 % WR Level I. 00 Actual Allowable Value

-- 15.0 % WR Level 00 i Minimum Allowable Value 14.87 % WR Level z

Z UJ Low Analytical Limit (AL) 13.50 %WR Level Note: The COT errors are based on the Minimum Trip Setpoint value minus the Minimum Allowable value and the Actual Trip Setpoint value minus the Actual Allowable Value.

Page 21

2.6 LOCA Alternate Source Termn Delaying the RS pump start will result in a short-term increase in air leakage from the containment and a short-term reduction in spray removal of radioactive isotopes from the containment atmosphere. To reflect the delay of the RS pump start, the following changes to the LOCA AST analysis are proposed:

1) Delay in RS operation for spray removal from 288.5 seconds to 40 minutes.

2) Spray volume for QS only operation, combined QS/RS operation, and RS only operation versus 1 sprayed volume for entire period of spray operation.

3) Early ORS pump start at 14 minutes for ECCS leakage vs. 288.5 seconds in the current basis.

4) RWST backleakage is assumed to start at 31.8 minutes vs. 30 minutes in the current basis.

5) Containment leakage after the first hour of a LOCA has increased to 0.04%-volume-per-day for the time period 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> vs. 0.02 1 %-volume-per-day for the time period 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in the current analysis.

6) Changes in aerosol removal coefficients due to the delay in RS operation and conservative QS flow rate assumptions.

7) Variable containment sump volume based on the containment analysis.

Other changes were made to the AST LOCA analysis to either remove conservative assumptions existing in the current analysis or changes based on a reanalysis of other parameters. These changes include:

1) Taking credit for the 96-hour to 720-hour control room occupancy factor listed in Regulatory Guide (RG) 1.183.

2) Taking credit for the timed release of nuclides into the containment sump in accordance with RG 1.183.

3) Increase the Decontamination Factor (DF) for releases from the RWST from 10 to 40.

4) For conservatism the containment volume has been increased to 1.9 16E+06 ft3 .

5) Increase the auxiliary building filter efficiency for organic iodines from 70% to 90% to be consistent with the Technical Specifications.

6) Increase the control room filter efficiency for organic iodines from 70% to 95% to be consistent with the Technical Specifications.

7) A slight increase in control room volume based on a recalculation.

8) The RWST "breathing rate" changed from 4 cfm to 3.7 cfmn.

Page 22

2.7 Containment Sump Surveillance Requirements Currently, the NAPS containment sump is shared by the SI and RS systems to perform the design functions for long-term core cooling and containment heat removal. The TS surveillance requirement (SR) for inspecting the containment sump is included only in the SI system (TS SR 3.5.2.8). However, the replacement sump strainer system will consist of separate strainers for the SI and RS systems. Therefore, it is proposed to add a surveillance requirement for an 18-month inspection of the RS system strainer (SR 3.6.7.7). As part of this change, TS SR 3.6.7.7 is being renumbered to TS SR 3.6.7.8 to be consistent with the Improved Technical Specifications practice of ordering surveillance requirements by frequency. In addition, SR 3.5.2.8 is changed to reflect the elimination of the existing trash racks and screens and to cover all of the SI sump components.

Page 23

3.0 GOTHIC Containment Analyses GOTHIC will replace the Stone & Webster LOCTIC code as the evaluation methodology in Chapter 6 of the NAPS UIFSAR for the containment design requirements described in Section 2. 1.

This section of the report documents two sets of GOTHIC analyses that were performed to demonstrate acceptable margins to the containment design criteria.

1. Current Configuration: The current RS system configuration with delay timers (400 seconds for IRS pumps, 210 seconds for ORS pumps), the current SI RMT automatic setpoint of 19.4% RWST level, the current TS 3.6.5 maximum air temperature of 120 F, and the current TS Figure 3.6.4-1 containment air partial pressure limits were used to demonstrate acceptable margins for the current plant configuration. These analyses are comparable to the LOCTIC analyses currently in the NAPS UJFSAR Chapter 6. GOTHIC margin improvements with respect to LOCTIC were described in Section 4 of topical report DOM-NAF-3. While some design inputs have changed from the LOCTIC analyses (see Section 3.1.4), transient behavior and results are similar to the LOCTIC UFSAR analyses. The current configuration analyses establish a set of baseline GOTHIC analysis results to which the proposed configuration cases are compared.

2. Proposed Configuration: The proposed configuration assumes four changes from the first set of analyses.

o3 The RS pumps are started assuming 60% RWST level coincident with a High High containment pressure signal. The ORS pumps start directly from the signal. The IRS pumps start 120 seconds after the actuation signal is reached. Instrument uncertainty is included for the level signal and the timer setpoint.

Li The SI RMT nominal setpoint is changed from 19.4% to 16% RWST WR level.

Instrument uncertainty is included for the level signal. The time to complete the RMT function ranges from 95-2 10 seconds (same as the current configuration analyses).

" The analyses that assume maximum containment temperature (see Table 3.11-2) revise the input from 121.5 F (current TS limit of 120 F + 1.5 F uncertainty) to 116.5 F (proposed TS limit of 115 F + 1.5 F uncertainty) for the containment air and the passive heat sinks.

o The containment air partial pressure is increased to the values in the proposed revision to TS Figure 3.6.4-1 that is provided in Section 3.10. The TS containment air partial pressure envelope is established to provide adequate margins to the containment acceptance criteria from the NAPS UFSAR, and to the proposed containment pressure limit of 2.0 psig from 1-6 hours after the LOCA per Section 4.

Page 24

The NAPS acceptance criteria for the proposed configuration containment analyses are:

" LOCA and MSLB containment peak pressure < 45 psig

" LOCA containment pressure < 2.0 psig from 1-6 hours and < 0.0 psig after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

" LOCA containment temperature < 280 F

" LHSI Pump NPSHa > Required NPSH

" ORS Pump NPSHa > Required NPSH

" IRS Pump NPSHa > Required NPSH Page 25

3.1 Application of the GOTHIC Methodology The GOTHIC analyses employ the containment response methodology described in topical report DOM-NAF-3 [3, 4, 17, 27] without modification. Benchmarks to LOCTIC containment response analyses from the Surry UFSAR were presented in Section 4 of DOM-NAF-3 and code differences in the treatment of key phenomena were described in that report. Benchmark analyses were performed for the NAPS GOTHIC models against eight LOCTIC analyses:

1) LOCA peak containment pressure (DEHLG break);

2) LOCA containment depressurization at 38 F SW (DEPSG break, minimum ESF);

3) LOCA containment depressurization at 75 F SW (DEPSG break, minimum ESF);

4) MSLB peak containment pressure (1.4 ft2 break at 0% power);

5) MSLB peak containment temperature (0.6 ft2 break at 102% power);

6) LHSI pump NPSHa pressure (DEPSG break with minimum ESF and maximum SI flow);

7) IRS pump NP`SHa (DEHLG break, 1 LHSI pump failure, maximum RWST temperature); and

8) ORS pump NPSHa (DEHLG break, 1 LHSI pump failure, minimum RWST temperature).

The containment response for each case was comparable to the Surry GOTHI1C analyses presented in References 3, 8 and 16. The differences in response with respect to LOCTIC are also the same as reported in DOM-NAF-3, Section 4. GOTIHIC predicts lower containment pressures and higher sump temperatures early in the accident. In addition, GOTHIC predicts faster containment depressurization times with margin in the post-QS peak pressure. The extensive benchmarking effort concluded that the North Anna GOTHIC models had been constructed appropriately and provided a good match to the LOCTIC (SWEC)JFROTH (Westinghouse) integral mass and energy releases to the containment for DEPSG and DEHLG breaks. Results from the benchmarking are not included in this report, because the current configuration analyses can be used to compare behavior against the LOCTIC analyses in the UIFSAR. The remaining part of this section reviews the key elements of the NAPS GOTHIC models used for the containment design analyses.

3.1.1 Model Geometry The NAPS containment is represented by a lumped control volume. The minimum and maximum free volumes are unchanged from the current analyses; the values are presented in Table 3.1-1. Control volumes are used to model the RWST and piping for the RS and SI systems. Junction elevations, heights, and loss coefficients are input consistent with the guidance in DOM-NAF-3, Section 3.2.1.

Nineteen thermnal conductors model the containment passive heat sinks for LOCA and MSLB analyses.

For LOCA analyses, flow paths model the break through the end of reflood using the vendor's mass and enthalpy data. At the end of reflood, the GOTH-IC simplified RCS model is activated. The release Page 26

from the first set of flow paths is stopped and different flow paths are activated from the RCS. For a DEPSG break, different flow paths model the release from the broken ioop cold leg and the broken loop pump suction during post-reflood. For a DEHLG break, different flow paths model the broken hot leg release from the vessel and the broken hot leg connection to the steam generator. A separate flow path and boundary condition inject the accumulator nitrogen into containment for LOCAs (Section 3.2.2 in DOM-NAF-3).

3.1.2 Engineered Safety Features The GOTHIC model includes a flow boundary condition to model the quench spray (QS) pumps.

Flow is variable as a function of the RWST level and downstream pressure. Pump heat is added when conservative. Pipe fill time and pump start delays are incorporated into a delay time that passes before the QS pumps deliver flow to the spray headers. A fraction of each QS pump flow is diverted to the suction of the respective IRS pump using boundary conditions.

Each RS pump is modeled with a flow boundary condition. Constant flow rates are assumed to bound the minimum and maximum delivered flow rates calculated from system analyses. RS pump heat is added when conservative. Trips are used to start the IRS and ORS pumps in accordance with the design description in Section 2.2. The trip delays include fill times for the RS pump discharge piping and time for the pumps to start and reach full flow. Control volumes are used for the RS pump suctions to allow the mixing of bleed injection flow and the accurate calculation of NPSHa at the pump first-stage impeller. Suction friction and form losses are included in the pump suction flow paths to accurately calculate NPSH available at the pump impeller. The casing cooling subsystem is modeled with a flow boundary condition for each pump that injects water to its respective ORS pump suction volume. The casing cooling boundary conditions are stopped with a trip when the available tank volume is exhausted.

Each of the four recirculation spray lines contains a single-pass, shell-and-tube heat exchanger located inside containment between the RS pump and the spray header. Heat exchanger performance is modeled to ensure a conservative prediction of heat removal from the sump for long-term accident analysis. The RSHX model selections in GOTHIC were benchmarked to a detailed heat exchanger design code over the range of accident flow rates and temperatures in the RS and SW systems. The HX models include tube plugging and fouling for analyses where it is conservative. Benchmark analyses demonstrated that the GOTHIC RSHX heat rates are comparable to LOCTIC after the containm-ent sump liquid temperatures converge to similar values.

Safety injection is modeled with flow boundary conditions that draw from the RWST and the containment sump. Before the end of reflood, sink boundary conditions remove mass from the RWST consistent with the vendor mass and energy calculation. At the end of reflood, the GOTHIC mass and energy model is activated and boundary conditions inject RWST water into the primary system. When Page 27

the RWST reaches a low-low level, the RWST boundary conditions are terminated and another boundary condition directs water from the containment sump to the primary system.

Nozzle components are used for each spray line. The Sauter mean diameter was calculated for each spray system in accordance with DOM-NAF-3, Section 3.4. 1. For containment integrity analyses, the nozzle spray flow fractions are set to 1.0 and the containment height is reduced using the methodology in Section,3.4.1.2 of DOM-NAF-3. The floor area gives the correct drop volume and surface area exposed to the containment atmosphere. For NPSH analyses, sensitivity studies showed that NPSHa is not sensitive to a reduction in containment height once the other assumptions that minimize NPSHa are implemented. Therefore, the containment height in the NPSH models is input from the containment free volume and the pool surface area.

3.1.3 Containment Passive Heat Sinks The containment heat sinks are grouped'into the following categories.

" Containment structure shell below grade

• Containment structure shell above grade

• Containment structure dome and liner

" Containment structure floor above floor liner

• Containment structure mat below floor liner

" Internal concrete slabs

" Carbon steel inside the containment

" Stainless steel inside the containment

" Accumulator tanks filled with water (MSLB only)

The DOM-NAF-3, Section 3.3, modeling guidelines for nodalizing thermal conductors were applied.

The surface area and thickness for concrete structures were taken from the current UFSAR analysis basis. The metal surface area and mass wvere increased from the current UFSAR minimum inventory based on a comprehensive review of containment metal that concluded that the previous inventory had omitted some structural metal and components. Thermal properties for concrete and steel were obtained from an engineering handbook, are presented in Table 3.1-2, and are the same as the Surry GOTHIC application [16]. Paint thermnal properties from the current LOCTIC analyses were confirmed to be conservative for the NAPS paint systems and were not changed. A contact resistance was modeled in the containment liner interface between concrete and carbon steel with a conductance of 40 Btulhr-ft2 -F, which is more conservative than the maximum value of 100 Btulhr-ft2 -F specified in DOM-NAF-3, Section 3.3. 1.

Heat transfer options were set consistent with DOM-NAF-3, Section 3.3.2. The Direct heat transfer option with DLM condensation was applied to all containment heat sinks except the sump Page 28

floor. The Split option was used for the floor to switch the heat transfer from vapor to liquid as the liquid level builds in the basement. The containment walls above grade and the containment dome used a specified external temperature of 95 F with a heat transfer coefficient of 2.0 Btulhr-ft2 -F. For NPSH analysis, a multiplier of 1.2 was applied to the Direct heat transfer coefficient.

In the LOCA and MSLB peak temperature cases, a 1 ft2 thermal conductor was added with the thickness of the containment liner and with a 1.2 multiplier on the DirectIDLM heat transfer coefficient to calculate a conservative containment liner temperature response. This is consistent with DOM-NAF-3, Section 3.3.3.

3.1.4 Plant Parameter Design Inputs During the development of the NAPS GOTHIC containment models, all of the containment analysis design inputs were reviewed and some values were revised. Key input changes from the current UFSAR analyses are summarized below. Table 3.1-1 summarizes the range of key input parameters from the NAPS GOTFHC containment analyses.

" The minimum surface area for metal heat sinks in containment was changed based on a revised inventory that was documented in an internal calculation. The passive heat sink data used in the GOTHI1C analyses is provided in Table 3.1-3. The metal and concrete heat sink minimum surface areas are 5% less than the nominal calculated values.

" Some of the assumed SI pump flow rates were revised based on hydraulic analyses of SI system performance. The range of assumed flow rates is listed in Table 3.1-1. The suction friction loss for one LHSI pump at maximum flow rate was revised from 9.2 ft to 8.8 ft using hydraulic analyses of the actual system configurations.

3.1.5 Containment Initial Conditions and -Instrument Uncertainty NAPS operates with a subatmospheric containment. As such, the selection of initial conditions for each accident analysis is consistent with Table 3.6-2 in DOM-NAF-3. The GOTHIC containment analyses include design inputs for plant parameters that are controlled by Technical Specifications, including containment air partial pressure, containment temperature, RWST temperature, and SW temperature. DOM-NAF-3, Section 3.6, describes how GOTHIC analyses could account for instrument uncertainty on the TS surveillance parameters in one of two ways. In the analyses in this report, instrument uncertainty was deterministically applied to the TS limit to develop a GOTHIC input (Option 1 in DOM-NAF-3). For example, the current TS limits on containment temperature are 86-120 F and the instrument uncertainty is 1.5 F. The GOTHIC analysis input range is 84.5-12 1.5 F. Table 3.1-1 defines the GOTHIC input assumptions for the TS parameters.

Page 29

3.1.6 NPSH Available and Water Holdup DOM-NAF-3, Section 3.8. 1, describes the licensing basis for calculation of NPSHa for the NAPS LHSI and RS pumps. A specific value for containment overpressure credit in the determination of NPSH has not been previously provided to the NRC for review and approval. Rather, NRC approval has been directed at verification of the adequacy of the methodology used to determine that the available NPSH is greater than the required NPSH for these pumps. The GOTHIC analysis methodology for NPSH in Section 3.8 of DOM-NAF-3 ensures that an overall conservative calculation is performed to minimize containment pressure and maximize containment sump temperature. DOM-NAF-3, Section 4.4 demonstrated the application of the conservative GOTHIC calculation of LHSI pump NPSHa for Surry Power Station, and the containment response compared favorably to the LOCTIC analysis of record. The same methodology was applied for NAPS in benchmark analyses and produced a comparable response to LOCTIC.

The NPSHa result from GOTHIC is based on the'conditions at the pump first-stage impeller elevation. The difference in elevation between the pump intake and the containment floor is included. Also, the pump suction friction and form losses (including the current clean sump screens) are specified in the junction between the containment and the pump. Therefore, the margin between the GOTHIC-calculated NPSHa and the required NPSH includes all essential elements of the problem except for strainer debris bed head loss, which is calculated external to GOTHIC and compared to the margin between NPSHa and required NPSH.

The NAPS NPSH calculations for the LHSI, IRS and ORS pumps employ the following conservative assumptions consistent with DOM-NAF-3, Section 3.8:

u A multiplier of 1.2 is applied to the DirectIDLM heat transfer coefficients for passive heat sinks.

o All of the spray water is injected as droplets into the containment atmosphere (nozzle spray flow fraction of 1) with the Sauter droplet size. Analyses are performed using the largest Sauter droplet size. A confirmatory ana lysis is performed by reducing the Sauter diameter by a factor of 2. The minimum NPSHa is repo rted from the case that provides the smaller NPSHa.

o The upper limit on containment free volume is used.

u The minimum containment air pressure is used.

o A minimum sump pool surface area is specified for the containment volume IJV interface area.

o For pump suction breaks, thermal equilibrium in the broken loop cold leg is forced using a liquid/vapor interface area of 1E+08 ft2 consistent with DOM-NAF-3, Section 3.5.3.3.2. This Page 30

promotes thermal equilibrium between any vapor from the downcomer and the SI added to that cold leg, which produces elevated sump temperatures. The SI flow is split between the downcomer (for the intact cold legs) and the broken loop cold leg using a flow distribution that is conservative compared to a hydraulic analysis of the RCS during a LOCA.

o A conservative water holdup volume is subtracted from the GOTHIC-calculated containment liquid volume to reduce the sump water height. Control variables incorporate the timing of spray system actuation and filling the refueling canal and calculate the total decrement to the GOTHIC containment liquid volume fraction. The corrected liquid volume fraction is then entered into a table of containment water level versus volume to determine the sump level to be used in the NPSHa calculation. For the LHSI pumps, the holdup areas in containment are filled before the pumps draw from the containment sump. The RS pumps start earlier in the accident and the holdup volumes are not filled completely at pump start. In the NAPS NPSH analyses, the containment holdup volume includes the following items:

1) water added to the RS and QS system piping,

2) water trapped from transport to the containment sump in the refueling canal and reactor cavity,

3) condensed films on heat structures,

4) films that form on platforms and equipment when spray is initiated, and

5) water absorbed in insulation.

The water level in the NPSH analyses is based on a planned modification to install a drain path between the reactor cavity and the outer containment basement. The NPSH analyses assume this drain for the calculation of the water holdup volume in the reactor cavity and in the determination of containment water level versus liquid volume.

Page 31

Table 3.1-1: Key Parameters in the Containment Analysis Parameter Value Maximum Core Power (102% x 2893 rated thermal power), MWt 2951 TS Containment Air Partial Pressure, psia TS Figure 3.6.4-1 (current)

______________________________________________ Figure 3.10-1_(proposed)

Containment Air Partial Pressure Uncertainty, psi +1-0.30 Containment Temperature, 'F (includes 1.5 'F uncertainty) 84.5 - 121.5 (current) 84.5 - 116.5 (proposed)

Containment Relative Humidity, % 0-100 SW Temperature, '~F (includes 3.0 'F uncertainty) 32-98 RWST Temperature, 'F (includes 2.0 'F uncertainty)' 32-52 Accumulator Pressure, psia 590-705 Accumulator Temperature, *F 84.5-121.5 Accumulator Water Volume, ft3 (includes uncertainty) 1007.3 - 1042.8 Accumulator Nitrogen Volume, ft3 407.2 - 442.7 Minimum Service Water Flow Rate with 2% RSHIX tube plugging, gpm 44102 Maximum Service Water Flow Rate with 0% RSHX tube plugging, gpm 9,000 LHSI Injection Mode Flow Rate (Single-Train), gpm 3066 -4201 Maximum LHSI Recirculation Mode Flow Rate (Single-Train), gpm 4050 HHSI Injection Mode Flow Rate (Single-Train), gpm 588-644 ORS Pump Flow Rate, gpm 3450 -3750 IRS Pump Flow Rate, gpm 3100 -3400 Minimum Casing Cooling Flow Rate to ORS Pump Suction, gpm 700 Casing Cooling Tank Available Volume, gallons 100,000 Casing Cooling Tank Maximum Temperature, 'F (includes 3.0 'F uncertainty) 53 Maximum Casing Cooling Delivery Delay from CDA signal, sec 55 QS Flow Rate, gpm Variable 3 QS Bleed Flow Rate to IRS Pump Suction, gpm 150 QS Spray Delivery Delay from CDA signal, sec 56 -70 LHSI Pump Suction Friction Loss at maximum 1-pump flow, ft 8.8 Page 32

Parameter Value ORS Pump Suction Friction Loss at maximum flow, ft 5.1 IRS Pump Suction Friction Loss at maximum flow, ft 0.42 CDA High High Containment Pressure, psia 30 RWST WR Level for RS Pump Start (60% +/- 2.5% uncertainty) 57.5% - 62.5%

ORS Pump Start Time Delay after 60% RWST level + CDA, seconds (0 or 10 0- 10 seconds for ramp to full flow depending on which is conservative)

ORS Piping Fill Time, seconds 46-6 1 IRS Pump Start Time Delay after 60% RWST level + CDA, seconds (+/- 12 108 -142 second timer uncertainty + 0 or 10 seconds for ramp to full flow, depending on which is conservative)

IRS Piping Fill Time, seconds 52-55 RWST WR Level Setpoint for RMT (Plant Setpoint +/- 2.5% uncertainty) 16.9 - 21.4% (current) 13.5 - 18.5 % (proposed)

Time to complete RMT function, seconds95-210 Minimum RWST volume at accident initiation, gallons 462,640 Current IRS Pump Start Delay, seconds 4 395 -405 Current ORS Pump Start Delay, seconds 4 205-215 Minimum containment free volume, ft3 1,825,000 Maximum containment free volume for NPSHa Analysis, ft3 1,916,000

1) Minimum RWST temperature of 32 F is assumed for evaluation of the inadvertent QS actuation event, but the GOTHIC analyses use 38 F. Normal operating range for RWST temperature is 40-50 F.

2) The minimum SW flow rate per RSHX is 4500 gpm with no tube plugging. The flow rate is reduced to 4410 gpm to account for 2% tube plugging.

3) The QS flow rate varies with the differential pressure between the containment (C) and RWST water level (L).

C-L, psid Minimum QS Pump Flow, gpm Maximum QS Pump Flow, gpm 52.1 1265.5 1465.5 39.94 1473.7 1673.7 27.77 1657.35 1857.35 20.61 1755.4 1955.4 13.44 1847.4 2047.4 4.12 1959.4 2159.4

-5.21 2066.8 2266.8

-13.01 2153.7 2353.7

-22.71 2255.3 2455.3

4) The current timer setpoints are used for "current configuration" analyses.

Page 33

Table 3.1-2: GOTHIC Model Heat Sink Material Properties Material Temperature Density Thermal Conductivity Specific Heat deg-F Ibm/ft3 Btulhr-ft-F Btullbm-F Carbon steel 70 490 27 0.10 Stainless steel 70 501 9.4 0.12 Concrete 75 142 1.0 0.156 Paint 75 110 0.125 0.10 Table 3.1-3: Containment Passive Heat Sinks TC # Description Minimum Surface Area, ft2 Thickness, inch 1 Interior Concrete Wall 1 7,741 6.006 2 Interior Concrete Wall 2 57,435 12.006 3 Interior Concrete Wall 3 51,064 18.006 4 Interior Concrete Wall 4 10,691 24.006 5 Interior Concrete Wall 5 8,674 27.006 6 Interior Concrete Wall 6 3,354 36.006 7 Cont Wall Below Grade 21,397 54.4026 8 Cont Wall Above Grade 28,090 54.4026 9 Containment Dome 24,925 30.5276 10 Containment Floor 11,757 146.699 11 Stainless Steel 0.3"-0.7" 9,378 0.360 12 Stainless Steel > 0.7" 330 1.490 13 Carbon Steel < 0.3" 74,920 0.225 14 Carbon Steel 0.3"-0.6" 12,304 0.333 15 Carbon Steel 0.6"!-1.0" 1,413 0.921 16 Carbon Steel 1.9"-2.0" 17,749 1.430 17 Carbon Steel > 2.0" 1,969 2.239 18 Galvanized Metal 95,667 0.069 19 Containment Liner Temperature Response 1 54.4026 20 EQ Conductor 1 1.0 29 More Carbon Steel 21,054 1.143 Page 34

3.2 Break Mass and Energy Release 3.2.1 LOCA Mass and Energy Releases The break release methodology in DOM-NAF-3, Section 3.5 is applied. The GOTHIC model assumes a constant drop size of 100 microns for the liquid release from the break until after the blowdown phase, at which time a continuous liquid is assumed. LOCA mass and energy release data through the end of reflood is obtained from the current licensing basis analysis from the NAPS steam generator replacement project in 1992. The NAPS mass and energy release analyses used the NRC-approved codes and methods documented in WCAP-8264-P-A [9] and WCAP-10325-P-A [10].

During the post-reflood phase, the GOTHI-C RCS system model is used to calculate the mass and energy release to the containment. The model was created using the guidelines in DOM-NAF-3, Section 3.5. The end-of-reflood mass and energy distribution in the primary system and steam generator secondary side is acquired from the Westinghouse mass and energy release analysis.

The mass and energy release accounts for the transfer of decay heat and the stored energy in the primary and secondary systems to the containment.

Lumped volumes are used for the vessel, downcomer, intact loop cold legs, broken loop cold leg, steam generator (SG) secondary side, up flow steam generator tubes and down flow steam generator tubes. Separate sets of loop and secondary system volumes are used for the intact and broken loops with the connections between the broken loop and containment as necessary for the modeled break location. Separate thermal conductors model the core, primary metal, SG tubes, and SG secondary metal.

The decay heat is modeled by specifying a time dependen 't internal heat generation for the fuel.

The 1979 ANS Decay Heat Standard is used consistent with DOM-NAF-3, Section 3.5.3.3.1.

The modeling approach outlined above successfully matched the long-term mass and energy release from the NRC-approved methodology employed in the current LOCTIC analysis of record

[8]. The GOTHIC simplified RCS model ensures that the stored energy in the core, primary metal, and the SG secondary has been released to the containment when the vessel is fully depressurized and the acceptance criteria for containment depressurization and NPSHa are challenged. Additional description about this modeling was provided in Response #8 to the NRC Request for Additional Information on the'Surry GOTHIC analysis submittal [8].

Page 35

3.2.2 MSLB Mass and Energy Releases The North Anna MSLB mass and energy data documented in WCAP- 11431 [11] was generated using the NRC-approved methodology from WCAP-8822 [12] and the LOF1TRAN computer code [13]. The analysis assumed two single failures: 1) the main steam non-return valve in the ruptured line fails to close, allowing all three SGs to blowdown until closure of main steam isolation valves; and 2) loss of one emergency bus. The NRC reviewed the MSLB mass and energy release data as part of a North Anna license amendment to increase the containment temperature limit. Section 5.3 in Attachment 3 of Reference 14 described the methodology and assumptions. The NRC Safety Evaluation Report was documented in Reference 15. The MSLB mass and energy release data from WCAP-1 1431 is applied in the GOTIHIC analyses in Section 3.7.

Page 36

3.3 LOCA Peak Pressure and Temperature The peak containment pressure is a function of the initial total pressure and average temperature of the containment atmosphere, the containment free volume, the passive heat sinks in the containment, and the rates of mass and energy released to the containment. The passive heat sinks in the containment are assumed to be at the same initial temperature as the initial average containment atmosphere temperature. Maximizing the initial containment total pressure and average atmospheric temperature maximizes the calculated peak pressure. The LOCA peak containment temperature is obtained from the peak pressure case because the containment atmosphere is saturated.

The double-ended hot leg guillotine (DEHLG) break causes a more limniting blowdown peak pressure than the double-ended pump suction break (DEPSG). The LOCA peak pressure analyses assume maximum initial containment pressure, maximum air temperature, 100% relative humidity, minimum containment free volume, and minimum heat sink surface area. Table 3.3-1 documents the results for GOTHIC calculations that use the current and proposed TS containment air partial pressure limits and containment temperature limits (120 F vs. 115 F). The change to the RS pump start method and the SI RMT setpoint do not affect the LOCA peak pressure and temperature because the peak values occur before the spray systems actuate and before SI RMT. The magnitude of the containment peak pressure is governed by the heat transfer to the containment passive heat sinks. For the proposed TS limits, the peak pressure increases 0.6 psi to 57.4 psia. This is the result of increasing the initial air partial pressure by 0.6 psi, which increases the initial air mass, which expands when heated from the break energy. The air pressure increase is offset somewhat by the reduction in initial containment temperature and corresponding reduction in vapor pressure. The colder heat sink surfaces condense slightly more steam during the blowdown.

For both cases, the containment peak pressure is less than the design limit of 59.7 psia. In addition, the containment vapor temperature and liner temperature remain below 280 F. Figures 3.3-1 and 3.3-2 compare the GOTHIC containment pressure and vapor temperature response for both DEI{LG cases.

The effect of RS pump operation is evident in the current configuration analyses after the outside RS pump delay time of 210 seconds from CDA passes.

The MSLB analysis in Section 3.7 produces a more limiting peak containment pressure than the LOCA analysis by 0.3 psi. LOCA and MSLB events set the TS containment air partial pressure maximum allowable value of 12.3 psia in Section 3.10. The maximum initial air partial pressure is independent of SW temperature, because the peak pressure occurs well before SW affects heat removal; therefore, the maximum allowable pressure is a constant line in Figure 3.10-1. The TS upper limit above 55 F SW in Figure 3.10-1 is limited by the containment depressurization analyses (see Section 3.4). In summary, a maximum operating containment air partial pressure of 12.3 psia ensures that the LOCA peak pressure is less than the design limit of 59.7 psia.

Page 37

Table 3.3-1: LOCA Peak Pressure and Temperature Analysis Results Current Proposed Configuration Configuration Initial Conditions TS Containment Air Partial Pressure, psia 11.7 12.3 Total Air Pressure (TS + 0.3 psi), psia 12.0 12.6 Initial Containment Temperature, F 121.5 116.5 Initial Containment Relative Humidity, % 100 100 Initial Vapor Pressure, psia 1.76 1.535 Initial Containment Pressure, psia 13.76 14.135 Results Peak Containment Pressure, psia 56.8 57.4 Time of Peak Containment Pressure, sec 19.12 19.07 Peak Containment Vapor Temperature, F 269.8 269.3 Peak Containment Liquid Temperature, F 251.0 250.5 Page 38

Figure 3.3-1: Comparison of Containment Pressure from DEHLG Peak Pressure Analysis 60-55-50-45-

.40 LD 35-30-25-20-15-10 -

0.1 10 100 1000 Time (sec)

Figure 3.3-2: Containment Vapor Temperature from DEHLG Peak Pressure Analysis 300 250 LL' a200 E

150 100 0.1 1 10 100 1000 Time (sec)

Page 39

3.4 LOCA Containment Depressurization The depressurization analysis is performed to show that the containment can be returned to subatmospheric conditions consistent with the assumption for containment leakage in the dose consequences analysis. Currently, the UFSAR depressurization analyses using LOCTIC show that the containment is subatmospheric within one hour and remains subatmospheric thereafter.

The current LOCA Alternate Source Term (AST) analysis assumes containment pressure is 0.5 psig from 1-4 hours and subatmospheric pressure after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> [20], but the margin after the first hour has not been used to relax the containment analysis yet. To support the proposed configu ration containment depressurization analyses, the AST licensing basis must be changed in accordance with Section 4.0 to accommodate a containment pressure less than 2.0 psig from 1-6 hours with subatmospheric pressure after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The time required to depressurize the containment and the capability to maintain it subatmospheric after a double-ended pump suction guillotine (DEPSG) break depends on the design of the containment depressurization systems, SW temperature, and the mass of air in the containment. The DEPSG break is limiting because it has the largest energy release to the containment due to the available energy removal from the SG secondary side. The loss of one emergency bus is the limiting single failure because it provides only one train of spray flow for containment atmosphere cooling. When SW temperature is elevated, it is more difficult to depressurize the containment and containment air partial pressure must be reduced to meet the depressurization limits.

Containment depressurization analyses were performed with GOTHIC for the current and proposed configurations to maximize the containment depressurization time (CDT) and the depressurization peak pressure (DPP). CDT represents the time when containment pressure first drops below atmospheric pressure. Once the operating QS pump is stopped after RWST depletion, only the RS system provides spray flow to the containment and at higher temperatures than the QS system (the maximum RWST temperature is 50 F). Once QS is terminated, the containment increases from subatmospheric conditions until it reaches the DPP, which is limited by the heat removal capacity of the RS system and the air mass in containment. A minimum initial containment temperature is conservative for DPP analyses, because higher initial air mass makes it more difficult to maintain subatmospheric conditions after QS termination. This response is evident in the current UFSAR anlasyes and in the GOTHIC analyses in this section.

Current Configuration The limiting case for the current configuration occurs for TS limits of 11.7 psia air partial pressure, 38 F SW, and 86 F air temperature. The containment response is very similar to the Surry benchmark analysis to LOCTIC in Section 4.4 of DOM-NAF-3. In the short-term accident response, GOTHIC predicts a lower peak pressure and a higher sump temperature based on the DirectfDLM condensation Page 40

and break effluent models. In the long-term, GOTHIC's lower containment pressure is attributed to the smaller superheated steam flow rate from the broken loop SG compared to the non-mechanistic Westinghouse FROTH analysis. The GOTHIC DEPSG model has removed the energy in the primary and secondary systems once the RCS is fully depressurized. .The integral mass and energy releases were very close to the LOCTIC analysis of record. Containment pressure reaches subatmospheric conditions at 2604 seconds and the DPP is -1.06 psig at 6077 seconds (QS terminates at 5841 seconds). The current configuration analyses maintain a subatmospheric containm-ent after one hour.

Proposed Configuration The proposed changes to the RS pump start method were incorporated and the containment air pressure was increased until sufficient margin was retained to the containment pressure limits imposed by the new LOCA AST analysis (i.e., 2.0 psig from 1-6 hours). The analyses are performed for SW temperatures from 55-95 F using the corresponding TS containment air partial pressure limits in Figure 3.10-1. Below 55 F, the MSLB and LOCA peak pressure analyses set the TS containment air partial pressure upper limit. The reduction in allowable containment air partial pressure as SW temperature increases is required to meet the LOCA depressurization. limits. The analyses in this section demonstrate that, for operation within Figure 3.10-1, the post-LOCA containment pressure is bounded by the pressure used to determnine the containment leakage in the LOCA dose consequences analysis in Section 4.

Several sensitivity analyses were performed for the proposed configuration to identify the most limiting CDT and DPP results. Table 3.4-1 summarizes the results from two final analyses that use the matrix of limiting assumptions specified in Table 3.11-2. Case 1 is the limiting analysis at the proposed TS maximum air partial pressure of 12.3 psia and 55 F SW. Case 1 pressure is less than 2.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, has a DPP of 0.78 psig in the 2nd hour after QS termination, and is subatmospheric within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Case 2 was analyzed at the proposed TS maximum air partial pressure of 10.4 psia at 95 F SW. Case 2 pressure decreased to less than 2.0 psig faster than Case 1 with a smaller DPP, but it takes longer to reach subatmospheric conditions because of the 95 F SW temperature. Containment pressure is subatmospheric in less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and remains subatmospheric thereafter.

Figures 3.4-1 (containment pressure), 3.4-2 (containment vapor and liquid temperature), and 3.4-3 (total RSHX heat rate) compare behavior from the current configuration case (TS limits of 11.7 psia, 86 F air, 38 F SW) and proposed configuration Case I (T'S limits of 12.3 psia, 86 F air, 55 F SW).

While the containment pressure and SW temperatures are different, the comparison illustrates the effect of delaying the RS pumps. As expected, the lack of RS spray and sump heat removal before 2180 seconds creates higher containment pressures and temperatures. However, the long-term pressures and temperatures are bounded by the assumptions in the LOCA AST analysis. The effect of the temperature and pressure profiles on equipment inside containment is addressed in Section 3.9.

Page 41

Small Break LOCA Containment Depressurization Changing the RS pump start from timers with fixed delays to an RWST level setpoint encouraged a review of the containment pressure and temperature response for small bre~ik LOCA (SBLOCA). The design basis large break LOCA causes a rapid pressurization of the containment and actuates a High High containment pressure signal within seconds. The RCS depressurizes quickly below the accumulator pressure and the LHSI pump shutoff head. The LHSI, HHSI, and QS pumps rapidly deplete the RWST inventory, such that the 60% RWST level setpoint is reached in a short period of time. The double-ended RCS pipe ruptures result in a large energy release to the containment and represent the most significant challenge to containment design criteria for peak pressure, peak temperature, and NPSHa for the LHSI and RS pumps. The large break LOCA analyses in this report have used the limiting single failures and ranged the possible pump flow rates to ensure that the most conservative response is obtained.

For SBLOCAs, the RCS pressure may stay above the LHSI pump shutoff head for a significant period of time. For this class of breaks, only the H!HSI pumps are available to feed the RCS. If the break is large enough to actuate a CDA, then the QS pumps will start and deplete the RWST.

However, until the RCS pressure is below the LHSI pump shutoff head, the RWST minimum depletion rate for one train of ESF (1 HHSI pump + 1 QS pump) is less than 3000 gpm. and the, time to reach the RS actuation setpoint of 60% RWST WR level is extended beyond the LBLOCA analyses in this section. Early in the event (-15 minutes), SBLOCA containment pressures and temperatures are lower than the LBLOCA response, but the SBLOCAs can extend the depressurization because of the slower drawdown of the RWST and release of RCS stored energy. As a result, the containment pressure and temperature profiles beyond 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> may be higher than those for RS pump start using the current timers, and the impact on EQ and the assumed containment leakage for dose conseq uences must be considered.

The NAPS SBLOCA Appendix K analysis for calculation of peak clad temperature (PCT) uses the Westinghouse NOTRUMIP computer code to determine the RCS response. The most recent analysis included 2", 3", 4" and 6" effective diameter cold leg break sizes [NAPS UFSAR Section 15.3. 1]. The 6" break produces a more rapid depressurization and accumulator actuation than the smaller breaks and PCT results are not reported in the UFSAR.. Break mass and energy release data was obtained from the NOTRUMP analyses for 3" and 6" break sizes. Breaks larger than 6" depressurize the RCS quickly, require LHSL flow early in the accident, And reach the RS actuation setpoint in a time frame approaching that of the LBLOCA cases, which have a bounding containment response. Breaks 3" and: smaller deplete the RWST slowly and would lead to procedure-driven operator action to depressurize the RCS using the secondary system.

The GOTHIC model for LOCA depressurization was used to predict the time to reach the RS pump start on 60% RWST level. Containment model assumptions were employed to maximize Page 42

the time to start the RS pumps (e.g., maximum initial RWST volume, RWST level setpoint of 57.5% for RS pump start, and minimum QS and SI flow rates) and maximize containment pressure and vapor temperature (1 train of ESF with minimum flow rates). Cases were analyzed along the minimum and maximum operating limits in the proposed TS Figure 3.6.4-1. The upper limit statepoints are 55 F SW and 12.3 psia air pressure and 95 F SW and 10.4 psia air pressure.

The lower limit statepoints are 10.3 psia air pressure over the SW temperature operating range.

The GOTHIC simplified RCS model was deactivated and NOTRUMP mass and enthalpy data was entered as a boundary condition to the containment. The data had to be extended to support the duration of the GOTHIC analyses. Starting at 2500 seconds for the 6" break, the break energy was based on the LBLOCA decay heat curve and conservative RCS stored energy. Energy removal through the SGs was ignored after 2500 seconds for conservatism. The energy release to the containment is conservative compared to expectations (the 3" break NOTRUMIP analysis showed primary-to-secondary heat transfer start at -1400 seconds). For the 3" break, the NOTRUMP mass and energy release data at 3000 seconds was extrapolated conservatively.

For the 6" break, the High High containment pressure was reached in less than 150 seconds and containment spray was delivered 70 seconds later. The RS pump start signal was reached at around 3700 seconds. Containment pressure at RS pump start was less than 29 psia and drops rapidly to subatmospheric conditions. For most cases, containment pressure was subatmospheric in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. For the most limiting case, containment pressure was subatmospheric in less than 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

For the 3" break, the High High containment pressure was reached at 410 seconds and containment spray was delivered 70 seconds later. The RS pump start signal was reached at around 5400 seconds. Containment pressure at RS pump start was less than 20 psia and drops rapidly once RS starts. The containment response for the 3" break was clearly bounded by the 6" break.

The purpose of the GOTHIC SBLOCA analyses was to show that the CDA actuation signal for QS initiation and the RS pump start signal would be reached to depressurize the containment to within acceptable limits for dose consequences and equipment qualification (EQ). The dose consequences from the SBLOCA are bounded by the LBLOCA. The SBLOCA containment pressure and temperature profiles were bounded by the EQ composite profiles. The GOTHIC analyses were conservative in assuming that long-term RCS energy was discharged only to the containment atmosphere. No credit was taken for energy removal from the SGs or for operator action to depressurize the RCS to use the LHSI pumps earlier and drain the RWST faster. For small LOCAs with a slow pressurization that takes a long time to reach the CDA setpoint, operators can initiate QS by procedure, which further ensures that containment pressure will be maintained within design limits for slow pressurization events.

Page 43

Table 3.4-1: Containment Depressurization Results for Proposed Configuration Case 1 Case 2 Initial Conditions*

TS Containment Air Partial Pressure, psia 12.3 10.4 Initial Containment Total Pressure, psia 13.2 11.3 TS Containment Air Temperature, F 86 86 TS SW Temperature, F 55 95 Event Time (seconds)

CDA High High containment pressure 2.6 3.2 SI flow initiated 27.0 27.0 Casing cooling flow delivered to containment 57.6 58.2 QS delivers spray to containment 72.6 73.2 End of reflood 253.4 253.4 ORS spray delivered to containment 2198.1 2177.0 IRS spray delivered to containment 2324.1 2303.0 Containment pressure < 2.0 psig 3205.0 3112.0 Switchover to SI recirculation mode complete 4442.4 4415.8 QS pump stopped 5857.2 5821.8 Depressurization peak pressure occurs 6680.0 7823.0

________________________________ (0.78 psig) (0.40 psig)

Casing cooling pump stopped 8626.9 8627.6 Containment pressure < 14.7 psia permanently 12,880 14,530

• Analyses include uncertainties of 0.30 psi air pressure, 1.5 F air temperature, and 3.0 F SW temperature.

Vapor pressure is 0.60 psia at 84.5 F (100% humidity).

Page 44

Figure 3.4-1: Comparison of Containment Pressure from DEPSG Depressurization Analysis 60 Curet oigurafion 55 PedCnfiguration Case 1 50 45 S40 35 30 2-5 20 15 10" 0.1 1 10 100 1000 10000 Time (sec)

Figure 3.4-2: Comparison of Containment Temperature from DEPSG Depressurization Analysis 300 250 S200 150 100 50 L.

0.1 10 100 1000 10000 Time (sec)

Page 45

Figure 3.4-3: Comparison of Total RSHX Heat Rate from DENSG Depressurization Analysis 500 450 Current Configuration

______ -- Proposed Configuration Case 1 4300 350

200 50 -______

0-0 1000 2000 3000 4000 5000 Time (sec)

Page 46

3.5 LHSI Pump NPSH Analysis A transient GOTHIC calculation is performed to demonstrate that the LHSI pumps have' adequate NPSH throughout the postulated LOCA. The NPSH available (NPSHa) must be greater than the NPSH required at all times during the accident. The difference between available and required NPSH is margin. The calculation of NPSHa with GOTHIC follows the methodology outlined in Section 3.8 of DOM-NAF-3. The DEPSG break provides the limiting LHSI pump NPSH results because it causes the largest energy release to the containment before RMT. Assumptions for key input parameters were based on the matrix of conservative assumptions for the Surry LHSI pump NPSH analysis from DOM-NAF-3, Section 4.7, and were confirmed with sensitivity studies. For the proposed configuration, the effect of delaying the RS pumps encouraged several sensitivity studies to be repeated.

The LHSI recirculation flow rate is conservatively assumed to be 4050 gpm based on one emergency bus as the most limiting single failure. This single failure leaves one LHSI and one HHSI pump, maximizes the pump suction friction loss, maximizes the LHSI pump required NPSH, and minimizes NPSHa. The analyses assume minimum heat sink surface area, minimum RS flow rates, minimum SW flow rate, maximum QS flow rate, maximum SI flow rates, and maximum containment temperature. The TS range for SW temperature (35-95 F) was analyzed with 3 F uncertainty.

Current Configuration Table 3.5-1 present's the LHSI pump NPSHa analysis results for the current configuration at 95 F and 73 F SW temperatures. The cases represent the* current limiting cases with LOCTIC for LHSI pump NPSHa for the current TS 'Figure 3.6.4- 1 that increases the containment air partial pressure limit from 8.85 psia at 73 F to 9.0 psia to 95 F. The LHSI pump minimum NPSHa of 14.49 ft occurs just after sump recirculation for a TS SW limit of 95 F. NPSHa increases to a value of 22.7 ft at 7200 seconds. High SW temperature is limiting because the RS pumps are removing sump energy for more than 2800 seconds before RMT is complete (see time sequence of events in Table 3.5-2). Higher SW temperature minimizes the containment energy removal during this long period of RS operation, although the effect of higher SW temperature is nearly offset by the lower air pressure at 73 F SW. Figures 3.5-1 (LHSI Pump NPSHa and water level),

3.5-2 (containment pressure and LHSI pump suction vapor pressure), 3.5-3 (containment vapor and liquid temperature), and 3.5-4 (RSHX heat rate) show the performance for the LHSI pump NPSHa analysis at 95 F SW.

Page 47

Proposed Configuration For the proposed configuration with delaying the RS pumps, the sensitivity studies on limiting single failure and plant parameters were confirmed. The results are summarized in Table 3.11-2.

Table 3.5-3 summarizes the LHSI pump NPSHa analysis results for the proposed configuration performed at 10 F SW temperature steps with a TS containment air partial pressure of 10.3 psia.

These analyses include the change to the SI RMT setpoint from 19.4% to 16.0% RWST WR level and the lower containment air temperature limit of 115 F. Table 3.5-4 provides the time sequence of events for select cases.

The delayed RS pump start reduces the system operating time before RMT from 2800 seconds to less than 1700 seconds, and NPSHa decreases. The increase in containment air pressure and lower system energy from the reduced heat sink initial temperature provide NPSH margin.

During this shorter window, lower SW temperature brings down the containment pressure quickly but the sump temperature holds up. In the current configuration, the maximum SW temperature was limiting, because the RS pumps start within 400 seconds of the CDA signal and had 2800 seconds of operation before RMT. Colder SW temperature would remove more energy from the containment through the RSHXs. In the proposed configuration, lower SW temperature has become limiting because the shorter operation period of RS before RMT provides less cooling of the sump liquid while still generating low containment pressures. There is a tradeoff between reduced spray temperature and reduced sump temperature. Once SW temperature drops below 75 F, the minimum NPSHa has little variability around 15.0 ft. As SW temperature decreases, both the containment pressure and sump temperature decrease and the effect of each change on NPSHa is offset.

Since the minimum NPSHa is observed to be stable over 10 F steps in SW temperatures, the selection of a limiting case is only necessary for graphing plots for the UFSAR. While several cases along the air partial pressure limit generate about the same minimum NPSHa, the analysis at 10.3 psia and 75 F SW temperature (Case 5) is selected as the limiting case for showing transient behavior. Figures 3.5-5 (LI{SI pump NPSHa and water level), 3.5-6 (containment and LHSI pump suction vapor pressure), 3.5-7 (containment vapor and liquid temperature), and 3.5-8 (RSHX heat rate) illustrate the performance of key variables for the LHSI pump NPSHa analysis at 75 F SW.

Note that Section 3.6 shows that the RS pumps have more NPSH margin than the LHSI pump for a containment air partial pressure of 10.3 psia. Therefore, the LHSI pump NPSH cases set the TS limit for minimum containment air partial pressure. '

Page 48

Table 3.5-1: LHSI Pump NPSHa Analysis Results - Current Configuration Initial Conditions Case 1 Case 2 TS Initial Containment Air Partial Pressure, psia 9.0 8.85 Initial Containment Total Pressure, psia 10.46 10.31 Initial Air Temperature, F 121.5 121.5 Relative Humidity, % 100 100 TS SW Temperature, F 95 73 Results at Time of Minimum NPSHa _______

Minimum NPSHa, ft 14.49 14.65 Margin to NPSH required of 13.82 ft 1.09 1.25 Time of minimum NPSHa, sec 3180 3168 Containment pressure, psia 10.53 9.69 Containment vapor pressure, psia 1.32 0.90 Containment liquid temperature, F 168.0 16 0.3 Containment vapor temperature, F 111.4 96.9 Water level, ft (referenced to 216.54 ft) 5.24 5.23 LHSI pump suction pressure, psia 11.48 10.63 LHSI pump suction vapor pressure, psia 5.73 4.80 Integral energy release, MBtu 745.8 745.8 Integral mass release, Mlbm 2.544 2.540 Table 3.5-2: Time Sequence of Events for LHSI Pump NPSHa Analysis (Current Configurmation_______

Time in seconds Case 1 Case 2 Accident Start 0 0 CDA Signal on High-High Pressure 4.1 4.2 Start SI 20.8 20.8 Casing cooling flow reaches containment 59.1 59.2 QS flow reaches containment 74.1 74.2 End of reflood phase 253.37 253.37 ORS flow reaches containment 263.7 263.7 IRS flow reaches containment 450.8 450.9 SI RMT initiated at 21.9% XVR level 3077.5 3066.1 SI RMT complete (95 second delay) 3172.5 3161.1 QS termination 5803.5 N/A Transient Termination 1 7200 1 3600 Page 49

Table 3.5-3: LUSI Pump NPSHa Analysis Results - Proposed Configuration Initial Conditions Case 1 Case 2 Case 3 Case 4 Case 5 Case 6 Case 7 TS Containment Air Pressure, psia* 10.30 10.30 10.30 10.30 10.30 10.30 10.30 TS SW Temperature, F 351 45 55 65 75 85 95-Containment Temperature, F 116.5 116.5 116.5 116.5 116.5 116.5 116.5 Results at Time of Minimum NPSHa Mfinimum LHSI Pump NPSHa, ft 15.13 15.03 15.11 15.01 14.97 15.19 15.48 Margin to NPSH required of 13.4 ft, ft 1.73 1.63 1.71 1.61 1.57 1.79 2.08 Time of minimum NPSHa, sec 3381 3383 j3384 3385 3388 j3390 3393 Containment total pressure, psia 10.53 10.72 10.98 11.21 11.43 11.79 12.28 Containment liquid temperature, F 166.7 168.5 170.2 172.1 173.7 175.6 177.7 Containment vapor temperature, F 79.1 84.5 91.1 96.4 101.2 108.4 116.8 Water level, ft (referenced to 216.54 ft) 5.46 5.46 5.46 5.47 5.47 5.47 5.47 LHSI suction pressure, psia 11.56 11.75 12.01 12.24 12.46 12.82 13.31 LHSI suction vapor pressure, psia 15.55 15.79 6.02 6.29 16.53 6.81 7.18 Integral energy release, UM tu (CV 20 1) 1761.0 1760.4 759.6 ,759.3 1757.3 756.7 754.7

• GOTIHIC total containment pressure is 11.535 psia from TS air pressure - 0.3 psi uncertainty + 1.535 psia vapor pressure for 100% humnidity at 116.5 F.

Table 3.5-4: Time Sequence of Events for LHSI Pump NPSHa Analyses - Proposed Configuration Time in seconds Case 1 Case 3 Case 5 Case 7 Accident Start 0.0 0.0 0.0 0.0 CDA Signal on High-High Pressure 3.6 3.6 3.6 3.6 Start SI 20.8 20.8 20.8 20.8 Casing cooling flow reaches containment 58.6 58.6 58.6 58.6 QS flow reaches containment 73.6 73.6 73.6 73.6 End of reflood phase 253.4 253.4 253.4 25.3.4 57.5% RWST level reached 1642.2 1642.2 1642.2 1642.2 ORS flow reaches containment (+71 seconds) 1713.2 1713.2 1713.2 1713.2 IRS flow reaches containment (+197 seconds) 1839.2 1839.2 1839.2 1839.2 Early SI RMT initiated at 18.5% RWST level 3278.1 3281.8 3284.8 3289.9 (16% setpoint + 2.5% uncertainty)

SI RMT complete (95 second delay) 3373.1 3376.8 3379.9 3384.9 QS termination 5570.9 5580.4 5590.9 5607.8 Transient Termination 7200 7200 7200 7200 Page 50

Figure 3.5-1: LUSI Pump NPSHa - Current Configuration (9.0 psia, 95 F)

Available NPSII LIISI Pump NPSII Available Analysis 25 -~LHSI Pump NPSH AvIailable Containment Water Level above 216.54 I

- NPSH Required ______

20 15.

z 10-0 -. . . . . . . . .

1000 2000 3000 4000 5000 6000 700D 8000 Time (sec)

Figure 3.5-2: Containment Pressure from LHSI Pump NPSHa Analysis - Current Configuration Containment Pressure LIISI Pump NPSII Available Analysis P!

I.

0.1 1 10 100 1000 10000 Time (sec)

Page 51

Figure 3.5-3: Containment Temperature from LHSI Pump NPSHa Analysis -

Current Configuration Containment Temperatures LIISI Pump NPSII Available Analysis S150 E

0.1 1 10 100 1000 10000 Time (sec)

Figure 3.5-4: Total RSHX Heat Rate from LHSI Pump NPSHa Analysis -

Current Configuration Total RSIIX Heat Rate LIISh Pump NPSII Available Analysis 300-250

-I-.

200-S U

150 S

4) 100-10 100 Time (sec) 10000 Page 52

Figure 3.5-5: LHSI Pump NPSHa - Proposed Configuration (10.3 psia, 75 F)

Available NPSII LuIST Pump NPSIT Available Analysis 3011 LHSI Pump NPSH Available Containment

.... Water Level above 216.54 I

-- NPSH Required 251 20.

z /0, 15-10.

5-0 1000 2000 3000 4000 6000 7000 8000

.Time (sec)

Figure 3.5-6: Containment Pressure from LHSI Pump NPSHa Analysis -

Proposed Configuration Containment Pressure LIISI Pump NPSII Available Analysis I.

0.

E 12 0.

0.1 1 10 100 1000 10000 Time (sec)

Page 53

Figure 3.5-7: Containment Temperature from LHSI Pump NPSHa Analysis -

Proposed Configuration Containment Temperatures LIISI Pump NPSII Available Analysis E

0.1 1 10 100 1000 10000 Timfe (sec)

Figure 3.5-8: Total RSHX Heat Rate from LHSI Pump NPSHa Analysis -

Proposed Configuration Total RSIIX hleat Rate LIISI Pump NPSII Available Analysis CD 1000 2000 3000 4000 5000 6000 7000 8000 Time (sec)

Page 54

3.6 RS Pump NPSH Analysis A transient GOT1HIC calculation is performed to demonstrate that the IRS and ORS pumps have adequate NPSH throughout the postulated LOCA. The NPSHa must be greater than the NPSH required at all times during the accident. The difference between available and required NPSH is margin. The calculation of NPSHa with GOTHIUC follows the methodology outlined in Section 3.8 of DOM-NAF-3. Section 3.7 demonstrates that the RS pumps are not needed for MSLB mitigation, so only LOCA events are analyzed for RS pump NPSHa.

Maximum RS pump flow rate is conservative for determining the NPSHa for that pump because it causes the highest suction friction loss and imposes that most restrictive NPSH required.

LOCTIC analyses have shown the IRS pump NPSH margin to be slightly more limiting than the ORS pump when considering the key factors that affect the pump suction conditions: 1) the ORS pump receives a minimum of 700 gpm casing cooling water (53 F) while the IRS pump receives 150 gpm of RWST water (52 F); 2) the IRS pump suction friction loss is 4.68 ft smaller (0.42 ft versus 5.1 ft for the ORS pump); 3) the ORS pump NPSH required is 1.7 ft higher (11.3 ft versus 9.6 ft) at the maximum pump flow rates; and 4) the IRS pump has 0.14 ft of extra head because the elevation of the pump impeller centerline is at -6.71 ft while the ORS pump impeller centerline is at -6.57 ft. With GOTHIC, the IRS and ORS pump NPSHa are tracked for all sensitivity runs to identify the specific break, single failure, and design inputs that minimize NPSHa for each RS pump.

Current Configuration For the current configuration, sensitivity studies were performed on break location, single failures, and key input parameters to develop a matrix of conservative assumptions as specified in DOM-NAF-3, Section 4.7. Tables 3.6-1 and 3.6-2 present the RS pump NPSHa analysis results for the current configuration for points along the current TS Figure 3.6.4-1. The DEHLG break is limiting because the mass and energy data maximize the energy in the containment sump early in the accident when the RS pumps are running. The limiting single failure is the loss of 1 LHSI pump (consistent with LOCTIC), which produced a slightly lower NPSHa than the assumption of no failure (full ESF). The TS statepoint of 8.85 psia and 73 F SW (Case 3) produces the limiting IRS pump NPSHa of 12.17 ft, with adequate margin to the NPSH required of 9.6 ft.

Figures 3.6-1 (available NPSH and water level), 3.6-2 (containment and IRS pump suction vapor pressure), 3.6-3 (containment vapor and liquid temperature) and 3.6-4 (RSHX heat rate) illustrate the performance of key variables from Case 3.

The minimum ORS pump NPSHa occurs for the same assumptions except for use of a minimum RWST temperature of 38 F (TS minimum 40 F - 2 F uncertainty) and minimum IRS pump flow.

Case 5 in Table 3.6-1 documents the results at the time of minimum ORS pump NPSHa and Figure 3.6-5 shows the transient NPSHa for the ORS pump. The minimum NPSHa of 15.30 ft has 4.0 ft of margin. The IRS pump is more limiting for the current configuration.

Page 55

Proposed Configuration For the proposed configuration that delays the RS pumps, sensitivity studies on break location, single failure and plant parameters were repeated. The results are summarized in Table 3.11-2.

Changing the RS pump start method causes a significant increase in NPSHa for the IRS and ORS pumps. While the containment pressure is much lower at the time of minimum NPSHa, the sump temperature is also much lower than the current configuration cases because the cold casing cooling water and QS bleed water is being injected directly to the sump before the RS pumps start. This extended cold water injection drops the average sump fluid temperature by a large

.amount in all cases. In addition, the delayed RS pump start allows for a higher water level to contribute to the NPSH margin gain. Based on these benefits, it is expected that the limiting failure would provide the smallest amount of.sump subcooling and water level benefit before RS system activation. It is seen later that the failure of a casing cooling pump and the loss of an emergency bus provide the smallest benefits to NPSHa. Eliminating casing cooling flow of 700 gpm at 53 F to the sump produces the lowest containment water level and hottest sump water.

Tables 3.6-3 and 3.6-4 compare the results of DEPSG and DEHLG analyses using the proposed plant configuration for several single failures. The analyses were performed at the proposed TS minimum air partial pressure of 10.3 psia and 35 F SW. The cold SW temperature produces cold RS spray that provides a fast containment depressurization. The DEPSG break produces a higher long-term energy release to the containment because of the available energy in the SG secondary side. Delaying the start of the RS pumps moves the pump operation into a time period when the DEPSG break energy produces a more limiting set of sump conditions. At the time of minimum NPSHa, the DEPSG case has a higher containment pressure, sump temperature, and water level.

The minimum NPSHa occurs later than the DEHLG case because it takes longer for the spray systems to depressurize the larger energy release and reduce the containment pressure. For all single failure scenarios with the delayed RS pump start, the DEPSG break produces a lower NPSHa than the DEHLG break.

The minimum NPSHa for the IRS pump is 15.12 ft (5.52 ft of margin) for the loss of an emergency bus (Case 2 in Table 3.6-3). The loss of QS bleed flow and the casing cooling pump on the failed bus provide the minimum subcooling and water level benefit. This more than offsets the lower spray flow rate compared to other cases. Figures 3.6-6 (available NPSH and water level), 3.6-7 (containment and IRS pump suction vapor pressure), 3.6-8 (containment vapor and liquid temperature), and 3.6-9 (RSHX heat rate) illustrate the performance of key variables for Case 2.

The minimum NPSHa for the ORS pump is 18.73 ft, with 7.43 ft of margin, for the loss of a casing cooling pump (Case 6 in Table 3.6-3). The IRS pump continues to have more NPSH margin than the ORS pump. Figures 3.6-10 through 3.6-13 show the behavior of key variables from the ORS pump NPSH limiting case.

Page 56

For the LOCA analyses in this section, the minimum containment water level is 1.86 ft above 216' 11" floor elevation (where the sump is located) when the ORS pump starts for Case 6 (loss of a casing cooling pump). This water level assumes a conservative holdup volume in containment of about 42,400 gallons and earliest pump start using 2.5% level uncertainty on the trip setpoint.

Long-term NPSH Margin for RS Pump When the RWST and casing cooling tanks are emptied, the injection flow to the RS pump suctions is lost. However, significant NPSH margin is available at these times and there is no adverse effect on long-term containment cooling. Case 6 was analyzed for 10,000 seconds to show the effects of exhausting the tanks. When QS stops on empty RWST at 2959 seconds, the IRS pump NPSHa changes from 32.1 ft to 31.7 ft but then continues to increase as sump temperature drops. When casing cooling stops at 8628 seconds, the ORS pump NPSHa changes from 31.6 ft to 28.8 ft and remains level for the duration of the analysis. The final containment water level is 7.0 ft above 216.54'. There is sufficient NPSH margin for long-term cooling from the RS system.

Page 57

Table 3.6-1: Results for RS Pump NPSHa Analyses (Current Configuration)

GOTHIC Case 4~ IRS IRS IRS IRS ORS Case 1 Case 2 Case 3 Case 4 Case 5 TS Containment Air Partial Pressure, psia 10.25 9.60 8.85 9.00 8.85 Initial Containment Pressure, psia* 11.71 11.06 10.31 10.46 10.31 Initial Air Temperature, F 121.5 121.5 121.5 121.5 121.5 TS SW temperature, F 35 55 73 95 73 Results Minimum NPSHa, ft 12.41 12.32 12.17 12.85 15.30 NPSH Required 9.6 9.6 9.6 9.6 11.3 Margin to NPSH Required, ft 2.81 2.72 2.57 3.25 4.00 Time of minimum NPSHa, sec 633.2 655.3 686.2 683.4 756.3 Containment pressure, psia 13.02 12.75 12.13 13.62 10.18 Containment liquid temperature, F 205.7 204.8 202.3 207.8 190.4 Containment vapor temperature, F 132.8 137.6 139.4 152.6 111.2 Water level, ft (referenced to 216.54 ft) 1.28 1.32 1.39 1.38 1.55 Pump suction pressure, psia 16.36 16.11 15.52 17.00 11.6 Pump suction liquid temperature, F 198.9 198.0 195.7 200.9 164.4

• GOTHIC total pressure is TS air pressure - 0.3 psi uncertainty + 1.76 psia vapor pressure.

Table 3.6-2: Time Sequence of Events for RS Pump NPSHa (Current Configuration)

Time in seconds IRS Case 3 ORS Case 5 Accident Start 0.0 0.0 CDA High High Pressure 4.19 4.19 Start SI 24.0 24.0 Casing cooling flow reaches containment 59.2 59.2 QS flow reaches containment 74.2 74.2 End of reflood phase 170.8 170.8 ORS flow delivered to containment 255.9 255.9 IRS flow delivered to containment 451.3 451.3 RS pump minimum NPSHa 686.2 (IRS) 756.3 (ORS)

Transient Termination 11800 11800 Page 58

Table 3.6-3: RS Pump NPSHa Results by Break and Single Failure for Proposed Configuration (10.3 psia, 35 F SW)**

GOTHIC Case-* 1 2 6 7 8 9 10 11 12 Break Location JEHG EPG DEHLG DEPSG DELILG DEPSG DEHLG DEPSG DEHLG DEPSG DEPSG DEPSG Single Failure Emergency bus QS Pump Casing Cooling Pump None LHSI Pump ORS pump IRS pump Results IRS Pump Minimum NPSHa, ft 19.77 15.12 20.94 17.71 17.62 15.78 18.85 17.57 18.50 16.81 118.28 17.76 Time of IRS minimum NPSI~a, sec 1895 2083 1623 1744 1306 1465 1263 1394 1313 1443 1478 1456

• Containmentpressure, psia 11.1 13.0 10.6 11.6 11.2 12.0 10.9 11.7 10.9 11.9 11.6 11.5 Containment liquid temperature, F 182.6 204.7 175.0 191.3 188.4 197.8 183.0 192.1 184.3 195.4 190.3 191.0 Pump suction liquid temp, F 176.8 197.8 169.6 185.1 182.4 191.3 177.2 185.9 178.5 189.0 184.1 184.8 Water level, ft (referenced to 216.54 ft) 3.07 3.44 12.93 2.87 2.72 3.15 2.80 3.18 2.81 13.18 3.48 3.43 Integral energy release, MBtu 483.6 628.0 469.2 593.0 443.8 558.0 440.1 549.7 440.4 552.3 558.6 557.1 ORS Pump Minimum NPS14a, ft' 20.34 19.31 20.56 19.59 18.97 18.73 19.47 19.45 19.28 19.17 20.06 19.70 Time of ORS minimum NPSHa, sec 1964 2250 1657 1811 1354 1518 # 1297 1454 1351 1519 1521 1506

• The DEHLG break analyses for IRS and ORS pump failures were analyzed. Detailed results were omitted from the table because they were bounded by the DEPSG.

• • All cases analyzed at 116.5 F containment temperature.

1. At the time of minimum ORS pump NPSHa, the containment pressure is 11.2 psia, the ORS pump suction pressure is 13.24 psia, the containment liquid temperature is 193.5 F, the ORS pump suction liquid temperature is 166.8 F, and the water level is 3.26 ft above 216.54 ft.

Page 59

Table 3.6-4: Time Sequence of Events from Limiting RS Pump NPSHa Analyses (Proposed Configuration)

Time reported in seconds Case 2 for IRS Pump Case 6 for ORS Pump DEPSG, emergency bus DEPSG, casing cooling failure pump failure Accident Start 0.0 0.0 CDA High High Pressure 3.63 3.63 Start SI 20.8 20.8 Casing coojing flow reaches 58.6 58.6 containment QS flow reaches containment 73.6 73.6 End of reflood phase 253.4 253.4 RS pump start signal reached 1439.9 1020.0 (62.5% WVR RWST level) __________

ORS flow delivered to containment 1485.9 1066.0 IRS flow delivered to containment 1599.9 1180.0 RS pump minimum NPSHa 2083 (IRS) 1518 (ORS)

SI RM4T starts 3151.8 2345.7 QS flow stopped on empty RWST > 3600 2958.5 Casing cooling tank emptied > 3600 8628.0 Transient Termination 3600 10,000 Page 60

Figure 3.6-1: IRS Pump NPSHa - Current Configuration (8.85 psia, 73 F)

Available NPSII Inside RS Pump NPSII Available Analysis 40 7

-IRS NPSH-Available 35- ----- Contalnment Water Level above 216.54 It

-- NPSH Fegiired t:H__

30-25 z

2 20 _____

Q9 I

15

.0 ...

0 200 400 600 B00 1000 1200 1400 1600 1800 Time (sec)

Figure 3.6-2: Containment Pressure from IRS Pump NPSHa Analysis -

Current Configuration Containment Pressure Inside RS Pump NPSII Available Analysis 40 30 Q.

0.1 1 10 100 1000 10000 Time (sec)

Page 61

Figure 3.6-3: Containment Temperature from IRS Pump NPSHa Analysis -

Current Configuration Containment Temperature Inside RS Pump NPSII Available Analysis IL 0

0 E

0 I-0.1 1 10 100 1000 10000 Time (sec)

Figure 3.6-4: Total RSHX Heat Rate from IRS Pump NPSHa Analysis -

Current Configuration Total RSIIX Heat Rate Inside RS Pump NPSII Available Analysis 1000.

900-_____

800-700 ______

600.______

c0 500-0 200 400 G00 B00 1000 1200 1400 1600 18D0 Timne (sec)

Page 62

Figure 3.6-5: ORS Pump NPSHa - Current Configuration (8.85 psia, 73 F)

Available NPSII Outside RS Pump NPSII Available Analysis 60 L -

ORS NPSH Available Containment Water Level above 216.54 ft NPSH Required ____

40.

z0 0.

z 30-

.0 20.

10 1--........

0 200 400 600 B0o 1000 1200 1400 1600 1800 Tknle (se Page 63

Figure 3.6-6: IRS Pump NPSHa - Proposed Configuration (10.3 psia, 35 F)

Available NPSII Inside RS Pump NPSII Available Analysis

-IRS NPSH Available

.... Ccttainment

. Water Level above 216.54 ft

-- NPSH Required ___________

35-30-25-S 20-15-_____

10- -_ - _ -_ _ _ _ _ _ -_ -_ -___-- - -__ - - - -

. .. .. . _ . .I... . .

5 _.. _. _. _. _. . . .. . .

0.

0 400 800 1200 1600 2000 2400 2800 3200 3600 Time (sec)

Figure 3.6-7: Containment Pressure from IRS Pump NPSHa Analysis -

Proposed Configuration Containment Pressure Inside RS Pump NPSII Available Analysis 60-50-40-

.2.

2! 30-0 2!

a.

20- 0 10-0.1 I 10 100 1000 10000 Time (sec)

Page 64

Figure 3.6-8: Containment Temperature from IRS Pump NPSHa Analysis -

Proposed Configuration Containment Temperature Inside RS Pump NPSII Available Analysis c

E A

0.1 1 .10 100 1000 10000 Time (we)

Figure 3.6-9: Total RSHX Heat Rate from IRS Pump NPSHa Analysis -

Proposed Configuration Total RSIIX Ifeat Rate Inside RS Pump NPSII Available Analysis

-400

,300 0 400 800 1200 1600 2000 2400 2800 3200 3600 Time (sec)

Page 65

Figure 3.6-10: ORS Pump NPSHa - Proposed Configuration Available NPSH Outside RS Pump NPSII Available Analysis t I 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 Time (sec)

Figure 3.6-11: Containment Pressure from ORS Pump NPSHa Analysis -

Proposed Configuration Contaimnent Pressure Outside RS Pump NPSII Available Analysis 0.1 1 10 100 1000 10000 Time (sec)

Page 66

Figure 3.6-12: Containment Temperature from ORS Pump NPSHa Analysis-Proposed Configuration Containment Temperature Outside RS Pump NPSII Available Analysis S150 E

4) 0.1 1 10 100 1000 1 0000 Time (sec)

Figure 3.6-13: Total RSHX Heat Rate from ORS Pump NPSHa Analysis -

Proposed Configuration Total RSIIX Ileat Rate Outside RS Pump NPSII Available Analysis 400 41 0 1000 2000 3000 4000 5000 6000 7000 8000 0000 10000 Time (sec)

Page 67

3.7 MSLB Peak Pressure and Temperature NAPS UFSAR Section 6.2.1.3.1.2 includes LOCTIC analyses for the containment response to a MSLB event. The existing NAPS analysis of MSLB containment pressure and temperature response will be replaced wvith the GOTHIC calculations described in this section. GOTHIC analyses in this section do not credit the recirculation spray system. This precludes the need to demonstrate adequate RS pump performance during a MSLB event. The limiting single failure in the containment model is the loss of an emergency bus, leaving one QS pump available with minimum flow and maximum time to deliver spray to containment.

3.7.1 MSLB Peak Pressure Analysis For the current LOCTIC analyses in the UFSAR, the maximum containment peak pressure occurs for the 1.4 ft2 break at 0% power because it has the highest SG liquid mass and results in the largest mass release to the containment before the faulted SG dries out. The UFSAR. analyses assumed that the RS system started with timer delays after containment pressure exceeded the CDA setpoint. The GOTIHIC MSLB analyses do not credit RS system operation, and the limiting MSLB case changes because of the lack of RS spray flow during the accident. With the assumption of an emergency bus failure, one QS pump is the only means of reducing containment pressure until the AFW flow to the faulted SG is isolated at 30 minutes. The long-term containment pressure and temperature are dependent: on the boiloff rate from the maximum AFW flow rate and the capacity of the operating QS pump.

Eleven combinations of break size and core power were analyzed to identify the peak containment pressure for the current and proposed TS maximum air pressure limits. Table 3.7-1 compares the results of the four most limiting peak pressure cases. Figure 3.7-3 compares the transient pressure profiles for eleven cases for the proposed configuration. The maximum peak pressure of 57.65 psia occurs for Case 2 (1.4 ft2 DER at 30% power) at 1812 seconds, shortly after AFW is terminated when the remaining SG liquid mass has been boiled to the containment. Table 3.7-2 shows the time sequence of events and Figure 3.7-1 shows the containment pressure for Case 2 with the proposed TS air partial pressure limit of 12.3 psia. This limiting case is a change from the current UFSAR analyses that show the 1.4 ft2 DER at 0% power to be limiting. This change is a direct result of not crediting the RS system. For some cases, the, high AFW flow rate to the faulted SG produces a pressure peak when the last of the SG liquid inventory is boiled shortly after AFW isolation. The high AFW flows combined with the high initial mass in the SGs at low power result in the change in the limiting case by a small amount of pressure margin. This change in the limiting case without crediting 'the RS system was not observed in the Surry GOTHIC containment analyses due to plant differences in the AFW and spray systems. The Surry containment spray pump capacity is higher and the maximum AFW flow rate of 400 gpm (limited by cavitating venturis) to the faulted SG is less than the 900-gpmn assumption for NAPS [16].

Page 68

The MSLB3 peak pressure of 57.65 psia is 0.3 psi higher than the LOCA peak pressure in Section 3.3.

The maximum initial air partial pressure is independent of SW temperature, because the RS system is not assumed to operate. Therefore, the maximum allowable TS air partial pressure is a constant line on Figure 3.10-1 until the containment depressurization analyses limit the curve (see Section 3.4). In summary, a maximum operating containment air partial pressure of 12.3 psia ensures that the MSLB peak containment pressure will be less than the design limit of 59.7 psia.

3.7.2 MSLB Peak Temperature Analysis Nine cases were analyzed with the assumptions that maximize containment temperature during a MSLB: minimum air partial pressure, maximum containment air temperature, and 0% humidity.

Analyses were performed for the current TS Figure 3.6.4-1 air partial pressure limit of 8.85 psia and the proposed TS minimum air partial pressure limit of 10.3 psia (Section 3.10). Table 3.7-3 compares the peak temperatures and Figure 3.7-4 compares the containment temperature profiles.

The maximum peak temperature occurs for the 0.6 ft2 break at 102% power (Case 6) very early in the transient. This result is consistent with the current UTFSAR analyses.

The increase in air pressure causes an increase in containment peak pressure but reduces the containment peak temperature. Figure 3.7-2 shows the containment vapor temperature for the proposed configuration Case 6, where the containment temperature peaks at 31 seconds when the break flow is reduced suddenly by the isolation of the non-faulted SGs from the steam line header. The vapor temperature decrease starting at 73 seconds is driven by the delivery of quench spray to the atmosphere. Containment pressure drops rapidly once operator action terminates AFW to the faulted SG at 30 minutes.

Section 3.9 describes the evaluation for the GOTHIC-predicted superheat conditions on safety-related equipment inside containment. LOCA temperature transient results are used for post-accident equipment qualification as allowed by IE Bulletin 79-01B [18] and its supplements. to IE Bulletin 79-OIB states that for a PWR MSLB inside containment, "equipment qualified for a LOCA environment is considered qualified for a MSLB environment in plants with automatic spray systems not subject to disabling single componenit failures." The North Anna spray systems meet this condition.

The analyses included a 1 ft2 thermal conductor to deterrmine a conservative containment liner temperature response in accordance with Section 3.3.3 of DOM-NAF-3. The conductor used a 1.2 multiplier on the DirectJDLM heat transfer coefficient. The peak liner temperature for the proposed configuration (case 6) was 258 F, so the sustained superheat does not adversely affect the containment liner.

Page 69

Table 3.7-1: Results from MSLB Containment Peak Pressure Analyses Case-) 1 2 3 4 2

Steam Line Break Size, ft 1.4 1.4 0.707 0.4 Break Type DER DER Split DER Core Power, % of Rated 0 30 30 30 Current Configuration* ____ ____ _____ ____

Peak containment pressure, psia 56.63 57.84 57.59 57.26 Time of peak pressure, sec 214.6 1812 1814 1825 Proposed Configuration# ________________

Peak containment pressure, psia 56.88 57.65 57.38 57.08 Time of peak pressure, sec 214.5 1812 1814 1825 1

• Current Configuration GOTHIC pressure is 11.7 psia TS air pressure + 0.30 psi uncertainty + 1.76 psia vapor pressure

1. Proposed Configuration GOTHIC pressure is 12.3 psia TS air pressure + 0.30 psi uncertainty + 1.535 psia vapor pressure Table 3.7-2: Time Sequence of Events from MSLB P eak Pressure Analysis - Proposed Configuration Time reported in Seconds Case 2 Accident start 0 CDA High High containment pressure 7 Start SI 27 QS delivered to containment 77 Faulted SG dryout 456 AFW terminated 1800 Containment peak pressure 1812 Transient Termination 7200 Table 3.7-3: Results from MSLB Containment Peak Temperature Analyses Case 4 5 6 7 Steam Line Break Size, ft2 0.7 0.6 0.4 Break Type DER DER DER Core Power, % of Rated 102 .102 30 Current Configuration* ____________

Peak containment temperature, F 306.2 318.4 298.9 Time of peak temperature, F 26.3 30.6 50.6 Proposed Configuration ____________

Peak containment temperature, F 296.2 308.4 291.9 Time of peak temperature, sec 26.4 30.8 50.6

• Current Configuration GOTHIC pressure is 8.85 psia TS air pressure - 0.30 psi uncertainty

1. Proposed Configuration GOTHIC pressure is 10.3 psia TS air pressure - 0.30 psi uncertainty Page 70

Figure 3.7-1: Containment Pressure from 30% Power, 1.4 ft2 MSLB Peak Pressure Analysis

- Proposed Configuration Containment Total Pressure 60 55 50 45

9- 40

-35

9. 30 25 20 15 10 0 1000 2000 3000 4000 5000 6000 7000 Time (sec)

Figure 3.7-2: Containment Temperature from 102% Power, 0.6 ft2 MSLB Peak Temperature Analysis - Proposed Configuration Containment Vapor Temperature 350 340 320 300

-280 S260 S240 220 E

200 180 160 140 120 0 1000 2000 3000 Time (sec)

Page 71

Figure 3.7-3: Containment Pressure Comparison from MSLB Peak Pressure Analyses - Proposed Configuration Containment Total Pressure Proposed Limits Peak Pressure Cases 65 30 poe

____________-0% 1. . t

-102% 0.7 ft2 40~- 102% 0.64 ft2

-70% 0.681 ft2

-30% 1.4 ft2 a.30 -30% 0.707

-30% 0.4 25-0% 1.4 Mt 1------0% 0.3 ft2 20 15 10 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (sec)

Page 72

Figure 3.7-4: Comparison of MSLB Peak Temperature Analyses - Proposed Configuration Containment Vapor Temperature Proposed Limits Peak Temperature Cases 320

ý102% cowe 0.6 ft2l 300 -

280 -102% 1.4 ft2

-102% 0.7 f t2

-102% 0.645 ft2

-102% 0.6 ft2

___L_---

-70% 0.681 Mt CL I30% 0.707 Mt 0 -- 0%/o 1.4 ft2 240 ____ _____________ -0% 0.3 Ut 220 200 . .L L...

0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (sec)

Page 73

3.8 Inadvertent QS Actuation Event NPSH calculations use the TS minimum containment air pressure as a design input. Historically, this limit has been set to provide sufficient operating margin with respect to the upper limit while balancing NPSH margin. The TS minimum limit also is an input to the event analysis for an inadvertent actuation of the quench spray system. This event is analyzed to verify three criteria from North Anna LJFSAR Section 6.2.6.3. Different portions of the containment liner can withstand different minimum pressures on the inside of the liner, as follows:

1. The shell and dome plate liners are capable of withstanding an internal pressure as low as 3 psia.

2. That portion of the bottom mat liner that is covered by concrete (i.e., everywhere but the sump) can withstand an internal pressure of 5.5 psia.

3. The bottom mat liner where exposed (i.e., the sump) due to its configuration is capable of withstanding an internal pressure as low as 5.5 psia.

Section 3.10 describes how the TS lower limit on containment air partial pressure will be increased to a constant 10.3 psia. An application of the equation of state for an ideal gas (Charles' Law) used in North Anna UFSAR, Section 6.2'.6.3, is repeated for the proposed TS limits. Design inputs, analysis, and conclusions are presented.

Initial Conditions Minimum air partial pressure (PI) 10.0 psia TS limit of 10.3 psia - 0.3 psi uncertainty Maximum bulk air temperature (TI) 116.5 F TS limit of 115.0 F + 1.5 F uncertainty Mvinimum RWST temperature, (T2) 32 F Bounding minimum value Saturation pressure at T2 (P.0) 0.09 psia ASME Steam Tables at 32 F Using Charles' Law for the air partial pressure (temperatures converted to Rankine), the final pressure in containment is calculated:

Par=":T2pl+ pSat (T2 )

Ptoal-":Par (460+ 32)(1 0.3- 0.3) + 0.09 =8.62 psia

=P.T+ (460+116.5)

Conclusions For an inadvertent QS actuation starting at the TS minimum air partial pressure of 10.3 psia and TS maximum air temperature of 115 F, the containment liner meets the design criteria cited above without operator action to terminate QS.

Page 74

3.9 EQ Envelope Verification Delaying the RS pump start and operating at higher containment air pressures affects the LOCA and MSLB pressure and temperature profiles. In this report, GOTHI-C containment pressure and temperature profiles were generated for LOCA peak pressure (Section 3.3), MSLB peak pressure and temperature (Section 3.7), and LOCA depressurization (Section 3.4). The GOTHIC pressure and temperature profiles were not bounded by the existing environmental zone description equipment qualification (EQ) envelopes, which were based on past LOCA analysis results.

Composite pressure and temperature profiles were developed that bounded the LOCA and MSLB pressure and temperature profiles from GOTHIC. The composite profiles were compared to the test reports for all environmentally qualified equipment inside containment, and it was concluded that the environmentally qualified equipment inside containment are qualified for the GOTHIC LOCA and MSLB accident analysis profiles for pressure and temperature in this report.

In conclusion, the EQ status of equipment inside containment is not affected by the GOTHIC containment temperature and pressure profiles resulting from the proposed configuration to delay RS pump start using RWST Level Low, increasing the containment air partial pressure limits in accordance with Figure 3.10-1, decreasing the containment temperature limit, and changing the RWST Level Low Low setpoint and TS allowable values for RMT.

Page 75

3.10 Proposed TS Limits for Containment Air Partial Pressure vs. SW Temperature Sections 3.3 through 3.7 describe GOTIHIC containment analyses that support an increase to the TS operating domain for containment air partial pressure. This increase is possible because of the margin gain in accident peak pressure from using GOTHIC instead of LOCTIC, and because of the improved containment depressurization times with the GOTHIC. methodology. A proposed change to North Anna TS Figure 3.6.4-1 is provided as Figure 3.10-1. This operating domain maintains the current limits of 35-95 F for SW temperature but reduces the maximum containment air temperature from 120 F to 115 F. Allowable limits for containment air partial pressure are defined by the following restrictions:

" The LOCA containment depressurization analyses in Section 3.4 limit the maximum operating air partial pressure to 12.3 psia at 55 F. At this same pressure, the LOCA and MSLB peak pressure analyses show margin to the containment design limit of 45 psig. Thus, the TS limit is maintained constant at 12.3 psia from 35 F to 55 F SW temperature.

" The containment depressurization analyses in Section 3.4 set the TS upper limit from 12.3 psia at 55 F SW to 10.4 psia at 95 F SW. The allowable air pressure decreases with increasing SW temperature because it is more difficult to depressurize the containment at higher SW temperature. To meet subatmospheric requirements, the initial air partial pressure is limited to 10.4 psia at 95 F SW.

o The LHSI pump NPSH analyses set the lower limit on air partial pressure (the RS pumps use the same assumptions but have more NPSH margin). The proposed lower limit in Figure 3.10-1 ensures at least 1.5 ft of NPSH margin for the LHSI pump at recirculation mode transfer across the entire SW temperature range.

This operating domain accounts for 0.30 psi instrument uncertainty for air partial pressure. For example, the proposed configuration LOCA and MSLB peak pressure analyses assume an initial total pressure of 14.135 psia (12.30 psia TS maximum air pressure + 0.30 psi uncertainty + 1.535 psia vapor pressure at 116.5 F and 100% relative humidity).

Page 76

Figure 3.10-1: Containment Air Partial Pressure versus Service Water Temperature (Proposed TS Figure 3.6.4-1)

Figure 3.6.4-1: Containment Air Partila Pressure Versus Service Water Temperature Ranges:

Containment Temperature 86-1 15*F RWST Temperature S 50*F 13.0 12.5

.(35,12.3)(5,23

12. UNACCEPTABLE OPERATION (0

10.5 ACCEPTABLE OPERATION 0.0 (35,10-3)1 UNACCEPTABLE OPERATION (95,10.3)

ý5 40 45 50 55 60 65 70 75 80 85 90 95 100 Service Water Temperature (OF)

Page 77

3.11 Summary of Containment Analysis Results Table 3.1 1-1 summarizes the GOTHI-C containment analysis results for the current and proposed plant configurations. The results for the proposed configuration demonstrate that all containment analysis acceptance criteria are met for operation in the allowable region of Figure 3.10-1 starting the RS pumps on 60% RWST WR level coincident with High High containment pressure. Table 3.11-1 includes a LOCA containment pressure limit of 2.0 psig during 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 /> based on the revised LOCA AST analysis in Section 4 of this report. GOTHIC MSLB temperatures greater than 280 F do not adversely impact the operation of safety-related equipment inside containment. LOCA transient pressure and temperature profiles will continue to be used for post-accident equipment qualification (refer to the licensing basis in Section 3.7).

Section 4.7 of DOM-NAF-3 identified that the limiting direction of key GOTHIC inputs would be identified for each containment acceptance criterion. Table 3.11-2 documents the results of the North Anna sensitivity studies for the proposed configuration to start the RS pumps on 60%

RWST level.

North Anna TS 5.5.15, Containment Leakage Rate Testing Program, requires a Type A containment integrated leak test in accordance with 10 CER 50 Appendix J. The maximum integrated leakage rate is limited to 0.1% by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the calculated LOCA peak pressure, Pa. The most recent NAPS Type A tests initialized the containment pressure greater than 44.1 psig, the current LOCA peak containment pressure in the North Anna UFSAR and TS 5.5.15. The GOTHIC-calculated LOCA peak containment pressure is 58.4 psia (42.7 psig) for the proposed TS maximum operating air partial pressure of 12.3 psia and TS maximum containment temperature of 115 F. The GOTHIC LOCA peak pressure is less than the current UIFSAR result used in the test procedure; therefore, the implementation of the change to TS Figure 3.6.4-1 is bounded by the most recent Type A tests. The value for Pa in TS 5.5.15 will be revised from 44.1 psig to 42.7 psig to be consistent with the GOTHIC LOCA containment peak pressure analysis.

Page 78

Table 3.11-1: GOTHIC Containment Analysis Results Acceptance Criterion Design Limit Current Proposed

________________________Configuration Configuration LOCA Peak Pressure 59. 7 psia 56.8 psia 57.4 psia LOCA Peak Temperature 280 F 269.8 F 269.3 F MSLB Peak Pressure 59.7 psia 57.54 psia 57.65 psia MSLB Peak Temperature* 280 F 3 18.4 F 308.4 F Containment Depressurization Time < 2.0 psig at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2081 sec to 3205 sec to

< 2.0 psig# < 2.0 psig Depressurization Peak Pressure < 2.0 psig 1-6 hours -1.06 psig +0.78 psig LHSI Pump NPSH 13.4 ft at 4050 gpm 14.49 ft 14.97 ft IRS Pump NPSH 9.6 ft at 3400 gpm 12.17 ft 15.12 ft ORS Pump NPSH 11.3 ft at 3750 gpm 15.30 ft 18.73 ft

• Refer to Sections 3.7 and 3.9 for the disposition of superheat MSLB conditions.

1. This analysis is subatmospheric within 2604 seconds and remains subatmospheric thereafter.

Page 79

Table 3.11-2: Matrix of Conservative Inputs for North Anna GOTHIC Containment Analyses Note: This table is based on the proposed plant configuration to start the RS pumps on 60% RWST wide range level coincident with High-High containment pressure. Other plant modifications can change these results and require sensitivities be evaluated against the results in the table.

Table Key-(also refer to the List of Acronyms and Abbreviations)

Min= Assume the minimum Value for the range of the design input Max =Assume the maximum value for the range of the design input N/A =Not Applicable: the key analysis result occurs before this parameter becomes effective or the component is not part of the containment response (e.g., accumulator nitrogen does not discharge for MSLB).

N/S = Not Sensitive: the key analysis result is not sensitive to changes in this input parameter.

LOCA Peak MSLR Peak Containme~~nt LHSI Pump ORS 'Pump IRS Pump Pressure* Pressure/Tenip # Depressurization J NPSH NPSH I NPSH General Break Type DEHLG 1.4 ft2 DER for pressure DEPSG DEPSG DEPSG DEPSG 0.6 ft2 DER for temp # ________

Reactor Power 102% 30% for pressure 102% 102% 102% 102%

102% for temp # ________

Single Failure N/A emergency bus emergency bus emergency bus casing emergency

____________________________ _________cooling pump bus Containment Air Pressure Max Max / Min # Max Min Min Min Temperature Max Max Min Max Max Max Relative Humidity 100% 100% / 0% # 100% 100%. 100% 100%

Free Volume Min Min Min Max Max Max Heat Sink Surface Area Min Min Min Min Min Min Page 80

LOCA Peak 1 MSLB Peak Containment LUST Pump ORS Pump IRS Pump Pressure*__ Pressure/Temp, # [Depressurization NPSH NPSH NPSH Safety Injection 11115 Injection Flow Rate N/A N/S Mini Max Min Mini LHSI Injection Flow Rate N/A N/A Min Max Mini Mini LHSI Recirc Flow Rate N/A N/A Mini Max N/A N/A LUST Suction Piping Friction Loss N/A N/A N/A Max N/A N/A Accumulator Nitrogen Pressure N/A N/A Max Mini Mini Min Accumulator Nitrogen Volume N/A N/A Max Mini Mini Mini Accumulator Nitrogen Temperature N/A N/A Mini Max Max Max RWST Temperature N/A Max Max Max Max Max Initial RWST Level N/A N/S Max Mini Mini Mini SI Recirc Mode Transfer N/A N/A Late Early N/S N/S Quench Spray QS Flow Rate N/A Mini Mini Max Max Max QS Start Time N/A Max Max Mini Miii Mini Bleed Flow to IRS Pump Suction NIA N/S N/S Miii Mini Mini Page 81

LOCA Peak1 MSLB Peak f Containment LHSI Pump PumpISPin IRS Pressure* J Pressure/Temp # Depressurization NPSH NPSH NS Recirculation Spray RS Piping Volume N/A N/A Max Max Max Max IRS Flow Rate N/A N/A Min Min Min Max ORS Flow Rate N/A N/A Min Min Max Min RS Pump Start on RWST Level N/A N/A Late Late Early Early IRS Suction Loss N/A NIA NIS N/S Max Max ORS Suction Loss N/A N/A N/S N/S Max Max Casing Cooling Flow Rate N/A Min Min Min Min Min Casing Cooling Tank Temperature N/A Max Max Max Max Max Casing Cooling Start Time N/A Max Max Max Max Max Service Water SW Flow Rate SW Temperature HX Tube Plugging/Fouling

• LOCA peak pressure and temperature assumptions are the same since a saturated containment environment is maintained.

1. MSLB peak temperature occurs for small breaks and the spectrum is reviewed for any plant operating parameter changes. The peak temperature is obtained by using minimum air pressure and 0% humidity (peak pressure cases assume maximum air pressure and 100% humidity).

Page 82

4.0 Revised LOCA AST Analysis Delaying the RS pump start will result in a short-term increase in air leakage from the containment and a short-term reduction in spray removal of radioactive isotopes from the containment atmosphere. As discussed in Section 2.6, the following changes to the design basis LOCA AST analysis [References 20 and 24] that reflect the delay in RS pump start are proposed:

1) Delay in RS operation for spray re moval from 288.5 seconds to 40 minutes.

2) Spray volume for QS only operation, combined QSJRS operation, and RS only operation versus 1 sprayed volume for entire period of spray operation.

3) Early ORS pump start at 14 minutes for ECCS leakage vs. 288.5 seconds in the current basis.

4) RWST backleakage is assumed to start at 31.8 minutes vs. 30 minutes in the current basis.

5) Containment leakage after the first hour of the LOCA has increased to 0.04%-volume-per-day for the time period 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> vs. 0.021 %-volume-per-day for the time period 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in the current analysis.

6) Changes in aerosol removal coefficients due to the delay in RS operation and conservative QS flow rate assumptions.

7) Variable containment sump volume based on the containment analysis.

Other changes were made to the AST LOCA analysis to either remove conservative assumptions existing in the current analysis or changes based on a reanalysis of other parameters, including:

1) Taking credit for the 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> control room occupancy factor listed in Regulatory Guide (RG) 1.183.

2) Taking credit for the, timed release of nuclides into the containment sump in accordance with RG 1.183.

3) Increase the Decontamination Factor (DF) for releases from the RWST from 10 to 40.

4) For conservatism, increase the containment volume to 1.9 16E+06 ft3 .

5) Increase the auxiliary building filter efficiency for organic iodines from 70% to 90% to be consistent with the Technical Specifications.

6) Increase the control room filter efficiency for organic iodines from 70% to 95% to be consistent with the Technical Specifications.

7) A slight increase in control room volume based on a recalculation.

8) The RWST "breathing rate" changed from 4 cfm to 3.7 cfm.

Page 83

The remainder of the LOCA AST analysis is unchanged from Section 3.1 of Attachment 1 to Reference 24. The unchanged assumptions include the following:

1) The source term and core power are unchanged.

2) The EAB, LPZ, and control room XIQ's are unchanged.

3) The dose conversion factors are consistent with Federal Guidance Reports 11 and 12.

4) The core release fractions and phases are consistent with RG 1.183

5) The chemical form of the iodines released from the fuel and also found in the sump is consistent with RG 1.183.

6) The modeling of elemental iodine spray removal is unchanged (X = 10 hr 1 , cutoff at 2.33 hrs).

7) The off-site and control room breathing rates are consistent with RG 1.183.

8) The sump pH > 7 when RS is credited for iodine removal.

.9) No credit is taken for removal of organic iodines by sprays nor any forms of iodines by deposition.

10) The control room is isolated at t=0 hours post-LOCA from a safety injection (SI) signal.

11) The control room unfiltered inleakage rate is 250 cfm. This is supported by tracer gas testing, which resulted in 150 cfm inleakage in a non-pressurized alignment of the control room.

12) One emergency control room fan is aligned to pressurize the control room with 900 cfm of filtered outside air at 60 minutes after the control room is isolated. Unfiltered in leakage of 250 cfm remains constant even during pressurization.

13) Control room filtered recirculation flow is not credited.

14) No credit is taken for the MCR bottle air system.

15) The flash fraction for iodines in the ECCS leakage analysis is consistent with RG 1.183 at 10%.

16) Auxiliary Building filtration (PREACS) is aligned at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA, ensuring ECCS leakage into area serviced by this system is filtered at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> post-LOCA.

17) EGGS leakage was modeled using differeint scenarios. They ranged from up to 3,400 cc/hr of only unfiltered leakage, 34,400 cc/hr of only filtered leakage, and combinations of both. The limiting leakage scenario is 3,400 cc/hr of unfiltered EGGS leakage. The allowable leakage limits are one-half of the analysis amounts per RG 1.183.

18) RWST backleakage rate is 2,400 cc/hr. The limit will be one-half of the analysis amount per RG 1.183.

19) The PREAGS and control room filter efficiencies for aerosols and elemental iodines remained at 98% and 95% respectively.

4.1 Changes in Containment Pressure and Leakage Assumptions In the current AST LOCA analysis the containment leak rate is modeled for the first hour at the peak pressure technical specification leak rate of 0.1% of the containment volume per day in Page 84

accordance with RG 1.183. For the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the LOCA, the current design basis analysis models the leakage at 0.021 % volume per day assuming a pressure of 0.5 psig. Containment leakage is modeled as terminating at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> based on containment attaining subatmospheric conditions.

In the revised design basis models the containment leak rate for the first hour remains at 0.1% of the containment volume per day. After the first hour, the containment leakage is conservatively modeled at 0.04% volumes per day, assuming a pressure of 2.0 psig, with leakage terminating after the sixth hour. This pressure profile bound the LOCA depressurization analyses described in Section 3.4. Table 4.1-1 summarizes the current and proposed containment pressure and leakage assumptions.

Table 4.1-1: Containment Leak Rate Assumption _______

Current Current Current Revised Time Revised Revised Time Period Containment Containment Leak Period Containment Containment Leak Pressure Rate Assumption Pressure Rate Assumption Assumption _______ Assumption 0-1 hours Decreasing to 0. 1 M/ay 0-1 hours Decreasing to 0. 1 %Ioday

___ ___ 1_0.5 psig 1 _ _ _ __ _ _ 1 _ _ _ _ 12.0 psigI 1-4 hours 0.5 psig 0.021 %/day 1-6 hours 2.0 psig 0.04 %17/day

> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Subatmospheric 0.0 > 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ISubatmospheric0.

4.2 Description of Containment Volumes The containment free volume is 1.91613+06 ft3 and the cross sectional area at the operating deck is 1.247134 ft2 . The containment free volume is an increase over that value reported in Reference 24, which is conservative for this analysis. For the first 40 minutes, only QS is operating. From 40 minutes to 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> both QS and RS are operating and after 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> only R S is operating.

Table 4.2-1 lists the volumes for the sprayed and unsprayed regions based on these times. The current analysis assumes a constant 70% sprayed volume starting at 90 seconds. The mixing rate of 2 times the unsprayed volume per hour was adjusted to reflect the volumes and times listed in Table 4.2-1. For conservatism the times listed in Table 4.2.1 reflect a further delay of the start of RS and an earlier termination of QS than those determined in the containment analysis.

Table 4.2-1: Time Dependent Sprayed[Unspray d Containment Fractions Time Period Percent Volume Percent Volume Sprayed Sprayed (ft3) Unsprayed Unsprayed (ft3 )

73 seconds - 40 minutes 37.6% 7.20413+05 62.4% 1.196E+06 40 minutes - 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 83.8% 1.606E+06 16.2% 3.10413+05 1.5 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 73.1% 1.40113+06 26.9% 5.154E+05 Page 85

The RADTRAD computer code modeled the sprayed volume and unsprayed volume as variable volumes using the values listed in Table 4.2-1. However the source term fraction can only be modeled as a single value in RADTRAD. To accurately reflect the source term fraction released simultaneously into the varying sprayed and unsprayed volumes a weighted average model was used. This model takes into account the size of the sprayed region and the period of time that the coverage existed. Since the source term is released over 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA, the final time period of RS only coverage is from 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The weighted average is calculated based upon the total core release period of 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the 3 time periods of spray operation, and the percent sprayed regions during those periods.

Weighted Average (0.67 hrs. x 0.376) + (0.83 hrs. x 0.838) + (0.3 hrs. x 0.731) 1.8 hrs This methodology results in a source term fraction of 64.8% in the sprayed region. The source term fraction released into the sprayed volume is conservatively rounded down to 0.64. The remaining, or 0.36, is released into the unsprayed volume.

4.3 Changes in Containment Spray Removal Coefficients With the delay in the start of the RS pumps, new aerosol removal coefficients were calculated using the same methodology as described in Reference 24. In this analysis QS start time is 73 seconds and termination is conservatively assumed at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Both ORS and IRS are conservatively assumed to start at 40 minutes.

Table 4.3-1 presents the characteristics of the QS and RS systems. The QS flow rates in Table 4.3-1 represent a decrease from the values presented in Reference 24. The QS flow rates used are conservatively assumed at 100 gpm lower than the flow rate values determined from the GOTHIC LOCA depressurization analyses in Section 3.4.

Table 4.3-1: Spray System Characteristics QS IRS ORS Elevation 391'-10" & 393'-2" Elevation: 377'-10" Elevation: 376'-10" 1400 gpm (73 - 2000 sec) 3450 gpm @ 40 minutes 3 100 gpm @ 40 minutes 1500 gpm (2000 - 3000 sec) ____________

1600 gpm (3000 - 5000 sec) ____________

1500 gpm (5000 - 5400 sec) _________________________

Page 86

The current and revised spray aerosol removal coefficients are calculated using the equations given in NUREGICR-5966. 'table 4.3-2 contains the existing aerosol removal coefficients.

Revised aerosol removal coefficients were developed as discussed below.

Table 4.3-2: Current Combined QS and RS Aerosol Removal Coefficients' Time (hours) Removal Coefficient (hr41)

From To 2.50E-02 8.0113-02 3.7267E+00 8.01E-02 1.33E-01 1.0799E+01 1.3313-01 1.56E+00 1.6672E+01 1.5613+00 1.80E+00 1.252813+01 1.8013+00 1.87E+00 7.9863E+00 1.87E+00 1.97E+00 5.5782E+00 1.97E+00 2.33E+00 2.976813+00 2.33E+00 3.76E+00 1.6191E+00 3.76E+00 5.35E+00 1.446013+00 5.3513+00 6.9713+00 1.4239E+00 6.9713+00 8.59E+00 1.421113+00 8.59E+00 1.6112÷02 11.4207E+00

1. Table 3.1-5 of Reference 24 and Table 15.4-6 of the UFSAR NUREG/CR-5966 [Page 173] presents the following equations for aerosol removal rate for the le~ percentile level:

2 2 In (A-=09 )= 5.5750 + (0.94362)ln Q - (7.327E - 7)Q H*2 - (6.9821E - 3)Q 2 H + (3.555E - 6)Q H Fr _(n 0.8945 lr+ .89nf Qi 0.9) 0.90.94

=0m9 = [0. 1108 -(0.0020 1)log 10 where X is the removal rate, mr is the mass fraction remaining in the containment, H is the spray drop height, and Q is the spray water flux, calculated by dividing the spray flow rate by the wetted cross-sectional area of the sprayed portion of the containment. The wetted cross-sectional area is determined by multiplying the containment cross-sectional area (1.247 ft2) by the sprayed fraction (percent sprayed from Table 4.2-1 divided by 100). The first equation above is used to calculate the removal rate corresponding to a mass fraction of 0.9. Using this value into the second equation yields the removal for a given value of mass fraction. Since the removal rate is dependent on drop height and spray rate, the spray headers have different removal rates.

Page 87

The drop heights and spray flux are calculated using input from Tables 4.2-1 and 4.3-1. Spray flux is derived as follows:

Q = (Spray Flow (t)gpm) (0. 13368 ftOgal) / (1.247E4 ft2) / (sprayed fraction (t)) (30.48 crn~ft) /1(60 sec/mmi)

Table 4.3-1 presents the QS and RS system characteristics using the 291'-10" elevation of the operating deck to determine the drop height. To simplify the modeling of the QS headers, both the upper and lower headers are modeled at the elevation of the lowest header or 391'-10" resulting in a drop height of 3048 cm. The 4 RS headers are modeled at the average elevation of the RS headers or 377'-4" resulting in a drop height of 2606 cm. This is appropriate since 1 train of IRS/ORS operating together supply water to both elevations.

When QS and RS are operating together a weighted average, based on flow rates, of the different elevations are used to calculate the drop height. A high QS flow rate is used for conservatism.L IRS (H) = (377'-10") - (291'-10") =86 ft = 2621 cm ORS (H) = (376'-10") - (291'10") = 85 ft = 2591 cm H = QS (3048 cm)(1800 gpm) + IRS (2621 cm)(3450 gprn) + ORS (2591 cm)(3 100 gpm) 8350 gpm H = 2702 cm during QS and RS operation NUREG/CR-5966 [Page 170] recommends that for a volume with continuing source, the removal constant associated with a mass fraction of 0.9 be used until the time-dependent source terminates. Hence, the mass fraction is assumed to remain at 0.9 from the start of the sprays until the end of the early in-vessel release phase at 1.8 hr. After this phase, the removal rate is adjusted stepwise by varying the mass fraction. The duration of time, t, required to change from a mass fraction mi0 to mrij is determined using the following formula:

m,= nw t = In(m*dmfl)/X For example, as seen in Table 4.3-3, it takes 0.08 hr (1.80 to 1.88 hr) to reduce the iodine mass fraction from 0.9 to 0.5. During this time step, the removal rate is constant at 7.739 hf 1.

Table 4.3-3 lists the data and aerosol removal coefficients for QS only operation, QS and RS operating simultaneously, and RS only operation.

Page 88

Table 4.3-3: Aerosol Removal Coefficients

_____ Q H Removal Constant (hif 1) Time (hr)

Mf (cm/sec) (cm) Xm=9 IXdX=9I Xf From To Aerosol Quench Spray Only Removal Coefficients 9.OOE-01 2.028E-02 3048 5.83213+00 1.00E+00 5.832E+00 2.03E-02 5.5613-01 9.0013-01 2.172E-02 3048 6.167E+00 1.OOE+00 6.16713+00 5.5613-01 6.6713-01 Aerosol Quench and Recirc Spray Removal Coefficients 9.OOE-01 5.23113-02 2702 1.25613+01 1.OOE3+00 1.256E+01 6.67E-01 8.33E-01 9.OOE-01 5.29613-02 2702 1.267E+01 1.OOE+00 1.26713+01 8.3313-01 1.11 E+00 9.OOE-01 5.29613-02 2702 1.26713+01 1.OO13+00 1.26713+01 1.1113+00 1.39E+00 9.OOE-01 15.231E-021 2702 11.25613+01 1 .OOE+00 11.25613+011 1.39E+00 I1.50E+00 Aerosol Recirc Spray Only Removal Coefficients 9.OOE-01 4.8813-02 2606 1.21413+01 1.OOE+00 1.21413+01 1.50E+00 1.80E+00 5.OOE-01 4.8813-02 2606 1.21413+01 6.375E-01 7.73913+00 1.8013+00 1.88E+00 3.OOE-01 4.88E-02 2606 1.214E+01 4.453E-01 5.40613+00 1.8813+00 1.97E+00 1.OOE-01 4.8813-02 2606 1.21413+01 2.376E-01 2.885E+00 1.97E+00 2.35E+00 1.OOE-02 4.88E-02 2606 1.21413+01 1.29313-01 1.569E+00 2.35E+00 3.8213+00 1.OOE-03 4.88E-02 2606 1.214E+01 1.155E-01 1.402E+00 3.8213+00 5.4613+00 11.OO13-04 4.8813-02 12606 1.21413+01 1.13713-0 1 1.38013+00 15.4613+00 17.1313+00 In the RADTRAD computer runs, which are limited to 10 values for aerosol spray removal coefficients, the removal coefficient at 6.67E-01 hours, the assumed start of RS, will remain constant at 1.256E+01 per hour until 1.5E+00 hours, the assumed end of QS. This is a conservative assumption as it can be seen in Table 4.3-3 that the removal coefficient increases during this time period.

4.4 Changes in ECCS Leakage Assumptions The delay in RS actuation results in a delay of the start of ECCS leakage. An early RS start for ECCS leakage results in a more conservative dose. Therefore, RS is assumed to start at 14 minutes instead of the 40 minutes assumed in the containment release. In the current analysis, ECCS leakage starts at 288.5 seconds [24].

Another change to the ECCS leakage analysis is in the modeling the containment sump volume as a variable volume in RADTRAD based on the GOTHIC analysis. A lower sump volume results in a higher dose due to less dilution volume. Therefore, the volumes reported by GOTHIC were reduced by 10% for conservatism. Table 4.4-1 lists the sump volume versus time used in the ECCS leakage analysis.

Page 89

Table 4.4-1: Containment Sump Volume vs. Time Time (seconds) Sump Volume (ft3) 840 16,800 1500 25,700 1900 31,400 2500 39,900 3000 46,800 4000 60,000 5000 68,800 6000 73,200 8000 76,000 4.5 Changes in RWST Leakage Assumptions As with the ECCS leakage analysis the containment sump volume is modeled as a variable volume in RADTRAD based on the GOTHIC analysis. A lower sump volume results in a higher dose due to less dilution volume. Therefore the volumes listed in Table 4.4-1 are also used in the RWST backleakage analysis. The start of RWST backleakage, which is the start of RMT, is consistent with the GOTIHIC analysis at 31.8 minutes.

The breathing rate used in Reference 24 for the RWST release was 4 cfm, whereas the RWST breathing rate used in this analysis was 3.7 cf~m. The analysis that supports Reference 24 calculated 3.7 cfm as the RWST breathing rate but rounded it up to 4 cfm for conservatism. This analysis uses the actual value calculated of 3.7 cfm.

The partition coefficient (PC) applicable to the iodines in the RWST water is based upon information in Reference 21. For this application, the RWST was assumed to behave like a closed system for the establishment of equilibrium conditions between the water and air. This is appropriate during the cooldown phase when air that is drawn into the RWST inhibits the loss of airborne iodine. It is also appropriate during the heat-up phase as the change in air volume is small and any impact on equilibrium conditions is therefore minimal.

The critical factor in the magnitude of the partition coefficient (PC) for iodines is the total iodine concentration in the water. 'For this application it was first necessary to compute the iodine concentration in the RWST. The ORIGENARP routine of SCALE calculated the total quantity of iodine in the core at 19,110 grams, which was conservatively rounded to 20,000 grams. The fraction of iodines released during the LOCA is 0.4, resulting in the containment sump containing 8,000 grams of iodine. The volume of liquid in the sump at the start of RWST backleakage is 31,400 ft3 (Table 4.4-1) or 234,900 gallons. The analysis of PC conservatively ignores the Page 90

increasing sump volume, which results in a lower concentration of lodines in the sump. The analysis of the PC also conservatively ignores the timing of the release into the sump over the 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Therefore, the maximum iodine concentration in the sump is 34 mgrams/gallon. The backleakage rate of 2400 cc/hr remains unchanged from Reference 24. Total sump liquid transferred to the RWST over 30 days as a result of backleakage is 457 gallons, resulting in a total of 15,560 ingrains of iodine transferred to the RWST.

Adding the 457 gallons, as a result of backleakage, to the minimum volume in the RWST at the end of the injection phase (16,780 gallons) results in a total volume in the RWST at the end of the 30 days of 17,240 gallons. Because the volume of water in the RWST is free of iodine, the concentration of iodine in the RWST increases over time and the maximum occurs at the end of the 30 days. Therefore the maximum concentration of iodine in the RWST is 0.9 ingrains/gallon or approximately 0.3 ingrains/liter.

At the maximum iodine concentration in the RWST, the partition coefficient (PC) and Decontamination Factor (DF) will be at a minimum. For conservatism, the DF at the end of 30 days will be used for the entire RWST backleakage period. The PC corresponding to the iodine concentration of 0.3 ingrains/liter is taken from Reference 21. Using the top curve, the PC is approximately 6000. The DF is calculated using the equation from SRP 6.5.2 and is:

DF = 1 + (Vliquid / Vair) *PC where Vliquid and Vair are the volumes in the RWST between which the partitioning takes place The RWST liquid volume at the end of 30 days is 17,240 gallons. The maximum RWST capacity is 509,900 gallons, resulting in an air volume at the end of 30 days of 492,600 gallons. The resulting DF is 211. The analysis conservatively uses a DF of 40.

4.6 Changes in Control Room Occupancy Factors Credit is being taken for the control room occupancy factors listed in RG 1.183. In Reference 24, the occupancy factor for the 96-hour to 720-hour time period was conservatively assumed at 0.6.

In the new analysis, the occupancy factor for that time period is 0.4, consistent with RG 1.183.

4.7 Timing of Release Phases The ECCS and RWST analyses from Reference 24 assumed that 40% of the core iodines were instantaneously transported from the core to the containment sumnp. The new analysis takes credit for the timed release of iodines into the sump as allowed by RG 1.183. The timed release of iodines in the sump is modeled as 5% for the first 30 minutes and 35% for the next 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Page 91

4.8 Control Room and Auxiliary Building Filter Efficiency In Reference 24 the control room and auxiliary building filter efficiency for organic lodines was conservatively modeled at 70%. In the new analysis the organic filter efficiency for the control room is modeled at 95% and the auxiliary building at 90%. These values are consistent with the North Anna Technical Specifications.

4.9 Control Room Volume The North Anna control room volume used in the analysis was recalculated to be 7.910E+04 ft3 .

This volume is a slight increase from 7.70E+04 ft3, which was used in Reference 24, but still remains a conservative value. The volume only includes the control room and the battery rooms minus a 10% conservative assumption and does not include other areas of the control room envelope.

4.10 Revised Radiological Results Table 4. 10-1 presents the revised design basis LOCA radiological dose results with the changes in assumptions as described in Sections 4.1 through 4.9. The analysis results are less than the regulatory dose limits.

Table 4.10-1: Revised Design Basis LOCA Dose Results Control Room Exclusion Area Low Population (Rem TEDE) Boundary Zone

____________________(Rem TEDE) (Rem TEDE)

Total Dose Consequences including contributions from containment, 4.1 2.1 0.2 ECCS and RWST leakage 10 CER 50.67 dose limits 5 25 25 Page 92

5.0 Conclusions This technical report demonstrates that the proposed safety analysis acceptance criteria are satisfied for the plant licensing basis changes outlined in Section 2. The specific changes for North Anna Power Station are:

o3 Start ORS pumps on 60% RWST WR level coincident with Hfigh High containment pressure Li Start IRS pumps on 60% RWST WR level coincident with High High containment pressure plus 120-second delay time o Incorporate the instrumentation and surveillance requirements for the RWST Level Low ESF function for RS pump start into the Technical Specifications.

" Replace the containment air partial pressure operating limits in TS Figure 3.6.4-1 with Figure

3. 10-1 in this report.

o Reduce the containment temperature operating limit in TS 3.6.4 and TS 3.6.5 from 120 F to 115 F.

o Change the TS allowable values for SI RMT to < 17.0% and > 15.0% in TS 3.3.2.

" Replace the LOCTIC containment analysis methodology in NAPS T.JFSAR Chapter 6 with the GOTHIC analysis methodology from the NRC-approved topical report DOM-NAF-3.

" Change the LOCA AST licensing bases as documented in Sections 2.6 and 4.

" Revise the containment pressure limit from 0.5 psig during the time interval from 1 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the LOCA initiation to 2.0 psig during the time interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the LOCA initiation.

The containment analyses were performned with the GOTHIC analytical methodology described in topical report DOM-NAF-3 [31, which was submitted to the NRC for generic review and approval on November 1, 2005 [4], supplemented in a letter dated July 14, 2006 [17], and approved by the NRC in a Safety Evaluation Report dated August 30, 2006 [27]. For North Anna Power Station, implementation of the GOTHIC methodology and analyses represents a change to a UFSAR evaluation methodology under 10 CFR 50.59 according to the screening performned in Attachment A.

Because of the changes assumed in this report, the analyses and license changes must be submitted to the NRC for approval. The containment analysis margins for the proposed plant changes are summarized in Section 3.11. Adequate margin to the acceptance criteria is demonstrated.

With the changes described in Section 2, the LOCA AST analysis results in Table 4. 10-1 show margin to the 10 CFR 50.67 limits for dose consequences. The AST analyses and revised technical bases must be submitted to the NRC for review and approval.

Page 93

6.0 References

1. NRC Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004.

2. Letter from David A. Christian (Dominion) to NRC, "Diominion Energy Kewaunee, Inc.,

Dominion Nuclear Connecticut, Inc., Virginia Electric and Power Company, Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2, Sunry Power Station Units 1 and 2, Response to NRC Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," serial number 05-2 12, September 1, 2005.

3. Topical Report DOM-NAF-3, Revision 0, "GOTHI-C Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment," October 2005.

4. Letter from Leslie N. Hartz (Dominion) to NRC, "~Dominion Energy Kewaunee, Inc. (DEK),

Dominion Nuclear Connecticut, Inc. (DNC), Virginia Electric and Power Company (Dominion),

Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Request for Approval of Topical Report DOM-NAF-3, GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment," serial number 05-745, November 1, 2005.

5. North Anna Power Station Updated Final Safety Analysis Report, Revision 41.

6. Technical Report NEI-04-07, Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Volumes 1 and 2 (Safety Evaluation Report), December 2004.

7. North Anna Power Station Technical Specifications.

8. Letter from Gerald T. Bischof (Dominion) to NRC, "Virginia Electric and Power Company, Sunry Power Station Units 1 and 2, Response to Request for Additional Information and Supplement to Proposed Technical Specification Change and Supporting Safety Analyses Revision to Address Generic Safety Issue 19 1," Serial No.06-545, July 28, 2006.

9. WCAP-8264-P-A, Revision 1, "Westinghouse Mass and Energy Release Data for Containment Design," August 1975. (WCAP-83 12-A is a Non-Proprietary version).

10. WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version," May 1983. (WCAP-10326-A is a Non-Proprietary version.)

Page 94

11. WCAP- 1143 1, Revision 0, "Mass and Energy Releases Following a Steam Line Rupture for North Anna Units 1 and 2," February 1987.

12. WCAP-8822-P, "Mass and Energy Releases Following a Steam Line Rupture," September 1976, with Supplements 1 (WCAP-8822-S 1-P-A) and 2 (WCAP-8822-S2-P-A) both dated September 1986. (WCAP-8860 is the Non-Proprietary version).

13. WCAP-7907-P-A, 'ILOFTRAN Code Description," April 1984.

14. Letter from W.L. Stewart (Virginia Power) to NRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Change," Serial No.

87-3 85, March 2, 198 8.

15. Letter from Leon B. Engle (NRC) to W. R. Cartwright (Virginia Power), "North Anna Units 1 and 2 -Issuance of Amendments Re: Containment Upper Limit Temperature (TAG Nos. 67535 and 67536)," December 14, 1988.

16. Letter from Leslie N. Hartz (Dominion) to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specification Change and Supporting Safety Analyses Revisions to Address Generic Safety Issue 191," Serial No.06-014, January 31, 2006.

17. Letter from Gerald T. Bischof (Dominion) to NRC, '¶Dominion Energy Kewaunee, Inc. (DEK),

Dominion Nuclear Connecticut, Inc. (DNC), Virginia Electric and Power Company (Dominion),

Kewaunee Power Station, Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Supplement to Request for Approval of Topical Report DOM-NAF-3, GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment," Serial No.06-544, July 14, 2006.

18. NRC LE Bulletin 79-01B3, Environmental Qualification of Class 1E Equipment, January 14, 1980.

19. ISA-RP67.04.02-2000, "Methodologies for the Determnination of Setpoints for Nuclear Safety Related Instrumentation,"

20. Letter from Stephen Monarque (NRC) to David A. Christian (Dominion), "North Anna Power Station, Units 1 and 2 -Issuance of Amendments on Implementation of Alternate Source Term (TAG Nos. MC0776 and MC0777)," June 15, 2005.

Page 95

21. "Iodine Removal From Containment Atmospheres by Boric Acid Spray," BNP-100, July 1970.

22. NUREG-CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays", D.A.

Powers and S.B. Burson, Sandia National Laboratories, Albuquerque, New Mexico, 1993.

23. Regulatory Guide 1.183, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.

24. Letter from Leslie N. Hartz (Dominion) to USNRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specification Changes Implementation of Alternate Source Term," Serial No.03-464, September 12, 2003.

25. NUREG-0800, Standard Review Plan Section 6.5.2, Revision 1 (1981), page C-10 and Revision 2 (1988), page 6.5.2-10.

26. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis for the RPS and ESFAS Test Times and Completion Times," October 1998.

27. Letter from Ho K. Nieh (USNRC) to David A. Christian (Dominion), "Kewaunee Power Station (Kewaunee), Millstone Power Station, Unit Nos. 2 and 3 (Millstone 2 and 3), North Anna Power Station, Unit Nos. 1 and 2 (North Anna 1 and 2), and Surry Power Station, Unit Nos. 1 and 2 (Surry 1 and 2) - Approval of Dominion's Topical Report DOM-NAF-3,

.'GOTHICMethodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment'," August 30, 2006.

Page 96

7.0 Regulatory Evaluation 7.1 No Significant Hazards Consideration The proposed changes to the North Anna Technical Specifications (TS) and licensing basis support the resolution of NRC Generic Safety Issue 191 (GSI-191), Assessment of Debris Accumulation on PWR Sump Performnance, and NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors. Seven changes are proposed as part of the requested amendment:

1) Replace the UFSAR evaluation methodology for analyzing the response to postulated pipe ruptures inside containment, including loss of coolant accident (LOCA) and main steam line break events, with the GOTHIC evaluation methodology in Dominion Topical Report DOM-NAF-3. The change to GOTHIC from the current LOCTIC code provides margins in LOCA peak containment pressure and other accident analysis results.

2) Increase the TS containment air partial pressure limits based on the GOTHIC containment analyses and LOCA Alternate Source Term (AST) analyses in Sections 3.0 and 4.0, respectively.

3) Reduce the TS maximum containment temperature limit from 120 F to 115 F.

4) Change the method of starting the recirculation spray (RS) pumps from timers, after a containment depressurization actuation on High High containment pressure, to refueling water storage tank (RWST) Level Low coincident with High H-igh containment pressure. This change ensures adequate water level to submerge the containment sump strainer and meets all safety analysis acceptance criteria. The proposed amendment modifies the North Anna Technical Specifications surveillance requirements to verify that each RS pump automatically starts on a CDA test signal after receipt of an RWST Level Low coincident with High High containment pressure. A plant modificati on associated with the proposed change to the Technical Specifications is required to install the new RS pump start circuitry.

5) Change the LOCA AST analysis basis to demonstrate acceptable dose consequences for the increased containment air partial pressure limits and the modification to RS pump start.

6) Change the TS allowable values for the safety injection (SI) automatic recirculation mode transfer (RMT) signal to be consistent with a plant setpoint change that is included in the GOTHIC containment analyses.

7) Modify the surveillance requirements for the sump to reflect the new containment sump strainer design.

Page 97

Dominion has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed changes to the North Anna Power Station Units 1 and 2 Technical Specifications and licensing basis and has determined that a significant hazards consideration does not exist. The basis for this determination is as follows:

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

No.

The proposed changes include a physical alteration to the RS system to start the inside and outside RS pumps on RWST Level Low coincident with High High containment pressure.

The RS system is used for accident mitigation only, and changes in the operation of the RS system cannot have an impact on the probability of an accident. The other changes do not affect equipment and are not accident initiators. The RWST Level Low instrumentation will comply with all applicable regulatory requirements and design criteria (e.g., train separation, redundancy, and single failure). Therefore, the design functions performed by the RS system are not changed.

Delaying the start of the RS pumps creates more challenging long-term containment pressure and temperature profiles. The environmental qualification of safety-related equipment inside containment was confirmed to be acceptable, and accident mitigation systems will continue to operate within design temperatures and pressures. Delaying the RS pump start reduces the emergency diesel generator loading early during a design basis accident, and staggering the RS pump start avoids overloading on each emergency bus. The reduction in iodine removal efficiency during the delay period is offset by changes to other assumptions in the LOCA dose analysis. The predicted offsite doses and control room doses following a design basis LOCA remain within regulatory limits.

The UPSAR safety analysis acceptance criteria continue to be met for the proposed changes to the RS pump start method, the proposed TS containment air partial pressure limits, the proposed TS containment temperature limit, the implementation of the GOTHIC containment analysis methodology, the proposed change to the SI RMT allowable values, and the changes to the LOCA dose consequences analyses. Based on this discussion, the proposed amendments do not increase the probability or consequence of an accident previously evaluated.

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

No.

Page 98

The proposed change alters the RS pump circuitry by initiating the start sequence with a new RWST Level Low signal instead of a timer after the High High containment pressure setpoint is reached. The timers for the inside RS pumps will be used to sequence pump starts and preclude diesel generator overloading. The RS pump function is not changed. The RWST Level Low instrumentation will be included as part of the Engineered Safety Features Actuation System (ESFAS) instrumentation in the North Anna TS and will be subject to the ESFAS surveillance requirements. The design of the RWST Level Low instrumentation complies with all applicable regulatory requirements and design criteria. The failure modes have been analyzed to ensure that the RWST Level Low circuitry can withstand a single active failure without affecting the RS system design functions. The RS system is an accident mitigation system only, so no new accident initiators are created.

The remaining changes to the containment analysis methodology, the containment air partial pressures, the maximum containment temperature operating limit, the TS allowable values for SI RMT, and the LOCA AST analysis basis do not impact plant equipment design or function.

Together, the changes assure that there is adequate margin available to meet the safety analysis criteria and that dose consequences are within regulatory limits. The proposed changes do not introduce failure modes, accident initiators, or malfunctions that would cause a new or different kind of accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously identified.

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

No.

The changes to the actuation of the RS pumps and the increased containment air partial pressure have created an adverse effect on the containment response analyses and the LOCA dose analysis. Analyses have been performned that show the containment design basis limits are satisfied and the post-LOCA offsite and control room doses meet the required criteria for the proposed changes to the containment analysis methodology, the RS pump start method, the TS containment air partial pressure limits, the TS containment temperature maximum limit, the TS allowable values for SI RMT, and the LOCA AST bases. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Page 99

7.2 Regulatory Requirements The regulatory requirements and standards applicable to the requested change are the following:

• 10 CER 50.49, Environmental Qualification Of Electrical Equipment Important To Safety For Nuclear Power Plants

• 10 CER 50.67, Alternate Source Term Sections 3.0 and 4.0 conclude that the proposed change will continue to comply with these regulatory requirements.

The GDC included in Appendix A to 10 CFR Part 50 did not become effective until May 21, 1971. The Construction Permits for North Anna Units I and 2 were issued prior to May 21, 1971; consequently, these units were not subject to GDC requirements. [Reference SECY-92-223 dated September 18, 1992] However, the plant was de signed to meet the intent of the draft GDC.

" Criterion 38--Containmient heat removal. "A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels."

There are no changes to the Recirculation Spray system or containment sump design that impact this general design criterion. Section 3.0 concludes that the proposed change will continue to comply with this regulatory requirement.

" Criterion 41--Containment atmosphere cleanup. "Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak. detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure."

Page 100

There are no changes to the Quench Spray and Recirculation Spray systems or containment sump design that impact this general design criterion. Section 3.0 concludes that the proposed change will continue to comply with this regulatory requirement.

• Criterion 50--Containmient design basis. 'The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any Joss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by

§50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters."

There are no changes to the Quench Spray and Recirculation Spray systems or containment design that impact this general design criterion. Section 3.0 concludes that the proposed change will continue to comply with this regulatory requirement.

I EEE-279 Standard,Nuclear Power Plant Protection Systems, August 1968.

The changes to the Recirculation Spray system actuation circuitry design meet this design standard. Section 2.0 concludes that the proposed change will continue to comply with design standard.

There are no changes to the Containment System or Quench and Recirculation Spray systems design or operation or the containment analysis method such that compliance with any of the above regulatory requirements and standards would come into question. The analysis completed to support the changes ensures the containment will continue to meet the applicable regulatory requirements. The plant will continue to comply with all applicable regulatory requirements.

In 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 operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Page 101

8.0 Environmental Assessment The proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CER 51 .22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described above, the proposed TS change does not involve a significant hazards consideration.

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

Delaying the start of the RS pumps creates more challenging long-term containment pressure and temperature profiles. However, the accident mitigation systems will continue to operate within design temperatures and pressures. The reduction in iodine removal efficiency during the delay period is offset by changes to other assumptions in the LOCA dose analysis and the dose consequences are within regulatory limits. The change to the TS allowable values for SI RMT has been included in the containment analyses that show margin to the design limits. The remaining changes to the containment analysis methodology, the containment air partial pressures, the containment temperature limit, and the LOCA AST analysis basis do not impact plant equipment design or function. Therefore, there is no significant change in the types or amount of any effluents that may~be released offsite.

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

Delaying the start of the RS pumps creates more challenging long-term containment pressure and temperature profiles. However, the accident mitigation systems will continue to operate within design temperatures and pressures. The reduction in iodine removal efficiency during the delay period is offset by changes to other assumptions in the LOCA dose analysis so that the dose consequences are within regulatory limits. The change to the TS allowable values for SI RMT has been included in the containment analyses that show margin to the design limits.

The remaining changes to the containment analysis methodology, the containment air partial pressures, and the LOCA AST analysis basis do not impact plant equipment design or function. Therefore, there is no significant change in the individual or cumulative occupational radiation exposure.

Based on the above assessment, Dominion concludes that the proposed change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CER 51.22 relative to requiring a specific environmental assessment or impact statement by the Commission.

Page 102

The proposed changes-the change to GOTHIC from the current LOCTIC code, the increase in the TS containment air partial pressure limits, the change to the method of starting the RS pumps from timers to the RWST level, the changes to the LOCA AST analysis basis, the change to the maximum operating containment temperature, and the change to the TS allowable values for SI RMT-provide additional margin to support the resolution of NRC GSI-191 and NRC Generic Letter 2004-02. The proposed changes have no adverse safety impact and do not significantly affect radiological dose consequences to the public or to plant workers.

Page 103

Serial No.06-849 Docket Nos. 50-338/339 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 MARKED-UP TECHNICAL SPECIFICATION PAGES VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

2. Containment SpraY5yX(: e.S

a. Manual Initiation 1, 2. 3. 4 2 per train, B SR 3.3.2.7 NA 2 trains

b. Automatic Actuation Logic 1, 2, 3, 4 2 trains C SR 3.3.2.2* NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

c. Containment Pressure High High 1. 2. 3 4 E SR 3.3.2.1 -ý28.45 psia SR 3.3.2.4 SR 3.3.2.8 SR 3.3.2.9

3. Containment Isolation

a. Phase A Isolation (1) Manual Initiation 1, 2. 3. 4 2 B SR 3.3.2.7 NA (2) Automatic Actuation 19 2, 3. 4 2 trains SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.5 (3) Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

b. Phase B Isolation (1)Manual Initiation Refer to Function 2.a (Containment Spray-Manual Initiation) for all functions and requirements.

(2) Automatic Actuation 1. 2, 3. 4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.5 (3) Containment Pressure High High Refer to Function 2.c (Containment Spray-Containment Pressure High High) for all functions and requirements.

North Anna Units 1 and-2 332Aedet 3.3.2-9 Amendments

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 4)

Engineered S-afety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER' SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

6. Auxiliary Feedwater

• a. Automatic Actuation Logic 1. 2. 3 2 trains G SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

• b. SG Water Level-Low Low 1, 2, 3 3 per SG D SR 3.3.2.1 Ž17%

SR 3.3.2.4 SR 3.3.2.8 SR 3.3.2.9

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

d. Loss of Offsite Power 1, 2, 3 1 per bus, F SR 3.3.2.6 Ž2184 V 2 buses SR 3.3.2.8 SR 3.3.2.9

e. Trip of all Main Feedwater 1. 2 2 per pump H SR 3.3.2.7 NA Pumps SR 3.3.2.9

7. Automatic Switchover to Containment Sump

a. Automatic Actuation Logic 1.2, 3, 4 2 trains C SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

b. Refueling Water Storage Tank 1, 2. 3, 4 4 I SR 3.3.2.1 a~n (RWST) Level-Low Low SR 3.3.2.4 :5zt~t*

SR 3.3.2.8 (/ )

SR 3.3.2.9 n-Coincident with Safety Refer to Functi on 1 (Safety Injectiion) for all initiation functions and Injection requirements.

8. ESFAS Interlocks

a. Reactor Trip, P-4 1, 2, 3 1 per train, F SR 3.3.2.7 NA 1:.

2 trains

b. Pressurizer Pressure. P-11 1, 2, 3 3 J SR 3.3.2.1 :52010 psig SR 3.3.2.8 C. Tavg-Low Low, P-12 1. 2. 3 1 per loop J SR 3.3.2.1 542OF and SR 3.3.2.8 :5 545'F North Anna Units 1 and 2332-1Aednt 3.3.2-11 Amendments CIfý 4j -

ECCS-Operati ng 3.5.2 SURVEILLANCEREQUIREMENTS_________

SURVEILLANCE FREQUENCY.

SR 3.5.2.5 Verify'each ECCS automatic valve in the 18 months flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.6 Verify each ECCS pump capable of starting 18 months automatically starts automatically on an actual or simulated actuation signal.

SR 3.5.2.7 Verify each ECCS throttle valve listed 18.months below is secured in the correct position.

Unit 1 Valve Number Unit 2 Valve Number 1-SI-188 2-SI-89 1-SI.-191 2-SI-97 1-SI-193 2-SI-103 1-SI-203 2-SI-116 1-SI-204 2-SI-111 1-SI-205 2-SI-12 SR 3.5.2.8 Verify, by visual inspection, Jac~h ECCS 18 months train containment sump 'uetion inkt- is not restricted by debris and the suti~R 1 ~inl&

tahiaLckz, anid st, t: show no evidence of structural distress or abn ~1 corrosion.

North Anna Units 1 and 2 3.5.2-3 3523Aedet Amendments "-""

ru-17 z'2ý-

Containment Pressure 3.6.4 Figure 3.6.4-1 (page 1 of 1)

Containment Air Partial Pressure Versus Service Water Temperature North Anna Units 1 and 23..-Amnet . 3.6.4-2 o.2 Amendment Nos. -2 4A

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be Ž: 86'F and

< ffff F APPLICABILITY: .MODES 1, 2, 3, and 4..

ACTIONS _________________________

CONDITION REQUIRED ACTION COMPLETION TIME A.. Containment average A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> air t'emperature not average air within limits, temperature to within limits.

B. Required Action and B.1 *Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCEREQUIREMENTS _________

SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify-containment average air temperature 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is within limits.

North Anna Units 1 and 2 3651Aedet 3.6.5-1 Amendments ZKýHý~-~-

RS System 3.6.7 SURVEILLANCE REQUIREMENTS_________

SURVEILLANCE FREQUENCY SR 3.6.7.6 Verify on an actual or simulated actuation 18 months

• signal(s):

.a.' Each RS automatic valve-'in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position; b..Each RS pump starts automatically; and

c. Each casing cooling pump starts automatically.

SR 3..71 Verify each spray nozzle is unobstructed. Following maintenance which could cause nozzle blockage 3 4 7,7 V_.t:',; 44 MlrCl /7-,

ýY,Mrtt -e-C.4 2- 8 M.OVA is O.al rwjAc_-_ 461 dewt/~s ~

North Anna Units 1 and 2 3673Aedet 3.6.7-3 ~f18a Amendments Mfti6-e-

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Safety Function Determination Program (SFDP) (continued) analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is.

inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the. system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a)and (b)above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCQ in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditi~ons and Required Actions to enter are those of the support system.

5.5.15 Containment Leakacge Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR.50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with'the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 as modified by the following exception:

NEI 94-01-1995, Section 9.2.3: The first Unit 1 Type A test performed after the April 3, 1993 Type A test shall be performed no later than April 2, 2008.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa', is044-.- psig. The containment design pressure is 45 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

'(continued)

.North Anna Units 1 and 25518Aednt 5.5-18 Amendments -2311/2 HE

TS Inserts 1 and A Insert #1: TS 3.3.2, Table 3.3.2-1 APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE CONDITIONS VALUE

d. Refueling Water 1,2,3 3 D SR 3.3.2.1 > 59% and Storage Tank SR 3.3.2.4 < 61%

(RWST) Level - SR 3.3.2.8 Low SR 3.3.2.9 Coincident with Containment Refer to Function 2.c (Containment Spray-Containment Pressure-High Pressure-High High High) for all functions and requirements.

INSERT A for TS 3.6.4: Replace Figure 3.6.4-1 with the below figure Figure 3.6.4-1: Containment Air Partial Pressure Versus Service Water Temperature Ranges:

Containment Temperature 86-11 5OF RWST Temperature -e,50OF 13.0 12.5I__

.~(35,12.3)

(55,12.3)

I 10.

E 12.0 0.

S11.5__ _ _ _ _

-(3,1.3 UACCEPTABLE OPERATION 10.5 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Service Water Temperature (OF)

Serial No.06-849 Docket Nos. 50-338/339 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 PROPOSED TECHNICAL SPECIFICATION PAGES VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

2. Containment Spray Systems I

a. Manual Initiation 1, 2, 3, 4 2 per train, B SR 3.3.2.7 NA 2 trains

b. Automatic Actuation Logic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

c. Containment Pressure High High 1, 2, 3 4 E SR 3.3.2.1 :528.45 psia SR 3.3.2.4 SR 3.3.2.8 SR 3.3.2.9

d. Refueling Water Storage Tank 1, 2, 3 3 D SR 3.3.2.1 59% and (RWST) Level-Low SR 3.3.2.4 :561%

SR 3.3.2.8 SR 3.3.2.9 Coincident with Containment Refer to Function 2.c (Containment Spray-Containment Pressure-High High)

Pressure-High High for all functions and requirements.

3. Containment Isolation

a. Phase A Isolation (1)Manual Initiation 1, 2, 3. 4 2 B SR 3.3.2.7 NA (2)Automatic Actuation 1, 2, 3. 4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.5 (3)Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

b. Phase B Isolation (1)Manual Initiation Refer to Function 2.a (Containment Spray-Manual Initiation) for all functions and requirements.

(2)Automatic Actuation 1, 2, 3. 4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.5 (3)Containment Pressure High High Refer to Function 2.c (Containment Spray-Containment Pressure High High) for all functions and requirements.

North Anna Units 1 and 2332- 3.3.2-9

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE

-MODESOR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

6. Auxiliary Feedwater

a. Automatic Actuation Logic 1, 2, 3 2 trains G SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

b. SG Water Level-Low Low 1,2.3 3 per SG D SR 3~3.2.1 2t17%

SR 3.3.2.4 SR 3.3.2.8 SR 3.3.2.9

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and I requirements.

d. Loss of Offslte Power 1, 2, 3 1 per bus, F SR 3.3.2.6 2: 2184 V 2 buses SR 3.3.2.8 SR 3.3.2.9

e. Trip of all Main Feedwater 1, 2 2 per pump H SR 3.3.2.7 NA Pumps SR 3.3.2.9

7. Automatic Switchover to Containment Sump

a. Automatic Actuation Logic 1,2, 3.4 2 trains C SR 3.3.2.2 NA and Actuation Relays SR 3.3.2.3 SR 3.3.2.5

b. RWST Level-Low Low 1,2 ,3.4 4 1 SR 3.3.2.1 Ž:15% and SR 3.3.2.4 :5 17% I SR 3.3.2.8 SR 3.3.2.9 Coincident with Safety Refer to Function 1 (Safety Injection) for all initiation functions and Injecti on requirements.

8. ESFAS Interlocks

a. Reactor Trip, P-4 1, 2, 3 1 per train, F SR 3.3.2.7 NA 2 trains

b. Pressurizer Pressure, P-11 1. 2, 3 3 J SR 3.3.2.1 :92010 psig SR 3.3.2.8 C. Tavg..Low Low, P-12 1.2, 3 1 per loop J SR :3.3.2.1 Ž5427F and SR 3.3.2.8 *545OF North Anna Units 1 and 2 3.3.2-11 3..21

ECCS-Operati ng 3.5.2 SURVEILLANCE REQUIREMENTS _________

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the 18 months flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal'.

SR 3.5.2.6 Verify each ECCS pump capable of starting 18 months automatically starts automatically on an actual or simulated actuation signal.

SR 3.5.2.7 Verify each ECCS throttle valve listed 18 months below is secured in the correct position.

Unit 1 Valve Number Unit 2 Valve Number 1-SI-188 2-SI-89 1.-SI-191 2-SI-97 1-SI-193 2-SI-103 1-SI-203 2-SI-116 1-SI-204 2-SI-1ll 1-SI-205 2-SI-123 SR 3.5.2.8 Verify, by visual inspection, each ECCS 18 months train containment sump component is not restricted by debris and shows no evidence of structural distress or abnormal corrosion.

North Anna Units 1 and 2352- 3.5.2-3

Containment Pressure

3.6.4 Ranges

Containment Temperature 86-1 157 RWST Temperature S 50*F 13.0 12.5

.(35,12.3)(5.23 UNACCEPTABLE OPERATION CL 0 ACCEPTABLE OPERATION S12.0

-(35,10.3) UNACCEPTABLE OPERATION (95,10.3) 35 40 45 50 55 60 65 70 75 60 85 90 95 100 Service Water Temperature (ff)

Figure 3.6.4-1 (page 1 of 1)

Containment Air Partial Pressure Versus Service Water Temperature North Anna Units 1 and 23.4- 3.6.4-2

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be Ž: 86*F and

• 5 115 0F. I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTI ONS _________________________

CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> air temperature not average air within limits, temperature to within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCEREQUIREMENTS _________

SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is within limits.

North Anna Units 1 and 2365- 3.6.5-1

RS System 3.6.7 SURVEILLANCE REQUI REMENTS SURVEILLANCE FREQUENCY SR 3.6.7.6 Verify on an actual or simulated actuation 18 months signal Cs):

a. Each RS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position;

b. Each RS pump starts automatically; and

c. Each casing cooling pump starts automatically.

SR 3.6.7.7 Verify, by visual inspection, each RS train 18 months containment sump component is not restricted by debris and shows no evidence of structural distress or abnormal corrosion.

SR 3.6.7.8 Verify each spray nozzle is unobstructed. Following I maintenance which could cause nozzle blockage North Anna Units 1 and 23673 3.6.7-3

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Safety Function Determination Program (SFDP) (continued) analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems. (a)and (b)above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.15 Containment Leakage Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 as modified by the following exception:

NEI 94-01-1995, Section 9.2.3: The first Unit 1 Type A test performed after the April 3, 1993 Type A test shall be performed no later than April 2, 2008.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42.7 psig. The containment design pressure *is45 psig.

c. The maximum allowable containment leakage rate, Laio at Pa2 shall be 0.1% of containment air weight per day.

(continued)

North Anna Units 1 and 2 551 5.5-18

Serial No.06-849 Docket Nos. 50-338/339 ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 MARKED-UP TECHNICAL SPECIFICATION BASES PAGES VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

TS Bases Inserts 2 through 14 Insert #2: Page B 3.3.2-14 When the RWST level reaches the low setpoint coincident with Containment Pressure-High High, the RS pumps receive a start signal. The outside RS pumps start immediately and the inside, RS pumps start after a 120-second delay. Water is drawn from the containment sump through heat exchangers and discharged to the RS nozzle headers.

Insert #3: Page B 3.3.2-14 RS is actuated manually or by RWST Level-Low coincident with Containment Pressure-High High.

Insert #4: Page B 3.3.2-16

d. RWST Level-Low Coincident with Containment Pressure-High High This signal starts the RS system to provide protection against a LOCA inside containment.

The Containment Pressure-High High (ESFAS Function 2.c) signal aligns the RS system for spray flow delivery (e.g., opens isolation valves) but does not start the RS pumps. The RWST Level-Low coincident with Containment Pressure-High High provides the automatic start signal for the inside RS and outside RS pumps. Once the coincidence trip is satisfied, the outside RS pumps start immediately and the inside RS pumps start after a 120-second delay.

The delay time is sufficient to avoid simultaneous starting of the RS pumps on the same emergency diesel generator. This ESFAS function ensures that adequate water inventory is present in the containment sump to meet the RS sump strainer functional requirements following a LOCA. The RS system is not required for SLB mitigation.

Automatic initiation of RS must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and sufficient energy exists in the primary and secondary systems to pose a threat to containment integrity due to overpressure conditions. The requirement for automatic initiation of RWST Level-Low to be operable in MODES 1, 2, and 3 is consistent with the operability requirements for Containment Pressure-High High. Manual initiation of the RS system is required in MODE 4, even though automatic initiation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA. In MODES 5 and 6, there is insufficient energy in the primary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond to mitigate the consequences of abnormal conditions by manually starting individual components. An operator can initiate RS at any time from the control room by using the pump control switch.

The manual function would be used only when adequate water inventory is present in the containment sump to meet the RS sump strainer functional requirements.

Insert #5: Pages B 3.6.4-1, B 3.6.6-1, B 3.6.6-2, B 3.6.7-1, B 3.6.7-2, B 3.6.7-3 to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Insert #6: Page B 3.6.4-1 Controlling containment air partial pressure limidts within prescribed limits ensures adequate net positive suction head (NPSH) for the recirculation spray and low head safety injection pumps following a DBA.

Insert #7: Page B 3.6.4-2 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.

Insert #8: Page B 3.6.4-3 the containment structure will depressurize to less than 2.0 psig in I hour and to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a DBA.

Insert #9: Page B 3.6.7-2 Refueling water storage tank (RWST) Level-Low coincident with Containment Pressure-High High provides the automatic start signal for the inside RS and outside RS pumps. Once the coincidence logic is satisfied, the outside RS pumps start immediately and the inside RS pumps start after a 120-second delay. The delay time is sufficient to avoid simultaneous starting of the RS pumps on the same emergency diesel generator. The coincident trip ensures that adequate water inventory is present in the containment sump to meet the RS sump strainer functional requirements following a loss of coolant accident (LOCA). The RS system is not required for steam line break (SLB) mitigation.

Insert #10: Page B 3.6.7-8 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.

Insert #11: Pages B 3.6.5-2 and B 3.6.7-3 The postulated SLB events are analyzed without credit for the RS system.

Insert #12: Page B 3.6.7-4 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.

Insert #13: Page B 3.6.7-8 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.

Insert #14: Page B 3.6.4-2 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.

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 1. Safety Injection (continued)

SAFETY.

ANALYSES, LCO, f. g. Safety Injection-High Steam Flow in Two Steam Lines_

AND *Coincident With Tav -Low Low or Coincident With Steam-APPLICABILITY Line Pressure-Low econtinued)

With the transmitters located inside the containment (Tavg) or near the steam lines (High Steam Flow), it is possible for them to experience adverse steady state environmental conditions during an SLB event.

The trip setpoint reflects only steady state instrument uncertainties.

This Function must be OPERABLE in MODES 1, 2, and 3 (above P-12) when a secondary side break or stuck open valve could result in the rapid depressurization of the steam line(s). This signal may be manually blocked by the operator when below the P-12 setpoint.

Above P-12, this Function is automatically unblocked.

This Function is not required OPERABLE below P-12 because the reactor is not critical, so steam line break is not a concern. SLB may be addressed by Containment Pressure High (inside containment) or by High Steam Flow in Two Steam Lines coincident with Steam Line Pressure-Low, for Steam Line Isolation, followed by High Differential Pressure Between Two Steam Lines, for- SI. This Function is not required to be OPERABLE in MODE 4, 5, or 6 because there is insufficient energy in the secondary side of the unit to cause an accident.

2. Containment Spray ýý'00) 5

~~ontainment Spray rovi deK t mar fun ctions:--

1. Lowers containment pressure antemperature after an HELB in containment;

2. Reduces the amount of radioactive iodine in the containment atmosphere; iwR4'

3. Adjusts the pH of the water in the containment sump after a large break LOCA; a.4t North Anna Units 1 and 2B332-3Rvso B 3.3.2-13 e Revision-e-

ESFAS Instrumentation B 3.3.2 APPLICABLE 2. Containment S ra 1(continued)

SAFETY ANALYSES, LCO, These functions are necessary to:

AND APPLICABILITY

• Ensure the pressure boundary integrity of the containment structure;

• Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure; and

• Minimize corrosion of the components and systems inside containment following a LOCA.

The containment spray actuation signal starts the .uesh ý7ý6 -

pmsand aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels con~tainment. Water is initially drawn from the R y thqene Spray pumps and mixed with a sodi* ydroxide solution from the chemical addition tank. hen the RWST reaches the low low level setpoint, the Low Head Safety Injection pump suctions are shifted to the containment sump. Containment spray is actuated manually or by Containment Pressure-High High signal.

a. Containment Spray-Manual Initiation CZ -C r The operator can initiate* containment spray at any time from the control room by simultaneously turning two containment spray actuation switches in the same train. Because an inadvertent actuation of containment spray could have such serious consequences, two switches must be turned simultaneously to initiate containment spray. There are two sets of two switches each in the control room.

Simultaneously turning the two switches in either set will actuate containment spray in both trains in the same manner as the automatic actuation signal. Two Manual- Initiation switches in each train are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function. Note that Manual Initiation of containment spray also actuates Phase B containment isolation.

North Anna Units 1 and 2 B3321 B 3.3.2-14 eis Rev nO-i si on

ESFAS Instrumentation B 3.3.2 ANALYSES, LCO, b. Containment Spray-Automatic Actuation Logic and AND Actuation Relays APPLICABILITY Automatic actuation logic and actuation relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b.

Manual and automatic initiation of containment spray must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and sufficient energy exists in the primary or secondary systems to pose.a threat to containment integrity due to overpressure conditions. Manual initiation is also required in MODE 4, even though automatic actuation i's not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA. However, because of the large number of components actuated on..a containment spray, actuation is simplified by the use of the manual actuation switches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system manual initiati~on. In MODES 5 and 6, there is insufficient energy in the pr iimary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond, to mitigate the consequences of'abnormal conditions by manually starting individual components.

c. Containment Spray-Containment Pressure This signal provides protection against a LOCA or an SLB inside containment. The transmitters Cd/p cells) are located outside of containment with the sensing line (high pressure side of the transmitter) located

.inside containment. The transmitters and electronics are located outside of containment. Thus, they will not experience any adverse environmental conditions and the Allowabl'e Value reflects only steady state instrument-uncertai nti es.

This is one of few Functions that requires the bistable output to energize to perform its required action. It is not desirable to have a loss of power (continued)

North Anna Units 1 and 2B332-5Rvso0B 3.3.2-15 Revision 0 .

• ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 2. Containment Spray jcontinued)

.SAFETY ANALYSES, LCO, c. Containment Spray-Containment Pressure (continued)

AND APPLICABILITY actuate containment spray, since the consequences of an inadvertent actuation of containment spray could be serious. Note that this-Function also has the inoperable channel placed in bypass rather thah trip to decrease the probability of an inadvertent actuation.

North Anna uses four channels in a two-out-of-four logic configuration and the Containment Pressure-High High Setpoint Actuates Containment Spray Systems.

Since containment pressure is not used for control, this arrangement exceeds the minimum redundancy requirements. Additional redundancy is warranted because this Function is energize to trip.

Containment Pressure-High High mutt be OPERABLE in MODES 1, 2, and-3 when there is sufficient energy in the primary and secondary sides to'pressurize the containment following a pipe break. In MODES 4, 5, and 6, there is insufficient energy in the primary and secondary sides to pressurize the containment and reach the Containment Pressure-High High setpoints.

3. Containment Isolation Containment Isolation provides isolation of the containment atmosp~here, and all process systems that penetrate containment, from the environment. This Function i's necessary to prevent or limit the release of radioactivity to the environment in the event of a large break LOCA.

There are two separate Containment Isolation signals, Phase A and Phase B. Phase A isolation isolates all automatically isolable process lines, except component cooling water (CC) and instrument air (IA), at a relatively low containment pressure indicative of primary or secondary system leaks. A list of the process

.lines is provided in the Technic'al Requirements Manual (Ref. 9). For these types of events, forced circulation cooling using the reactor coolant pumps (RCPs) and SGs is the preferred (but not required) methdd of decay heat removal. Since CC is required to support RCP operation, not isolating CC on the low pressure Phase A signal (continued)

North Anna Units 1 and 2 B 3.3.2'-16 B3321 eiin~

Revision ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 3. Containment Isolation (continued)

SAFETY ANALYSES, LCO, a. Containment Isolation-Phase A Isolation .(continued)

AND APPLICABILITY conditions and manually actuate individual isolation valves in response to abnormal or accident conditions.

(3)Phase A Isolation-Safety Injection Phase A Containment Isolation is also initiated by all Functions that initiate SI. The Phase A Containment Isolation requirements for these Functions are the same as th6 requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

b. Containment Isolation-Phase B Isolation Phase B Containment Isolati-on is accomplished by Manual Initiation, Automatic Actuation Logic and Actuation Relays, and by Containment Pressure channels (the same channels that actuate Containment

-. )_Spray, Function 2). The Containment Pressure trip of ase Containment Isolation is energized to trip in order to minimize the potential of spurious trips

• that may damage the RCPs.

(1) Phase B Isolation-Manual Initiation

• (2) Phase B Isolation-Automatic Actuation Logic and Actuation Relays Manual and automatic initiation of Phase B containment isolation must be OPERABLE in MODES .1,2, and 3, when there is a potential for an accident to occur. Manual initiation is also required in MODE 4 even though automatic actuation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA. However,

-because of the large number of components actuated-on a Phase B containment isolation, actuation is simplified by-the use of the Containment Spray manual actuation switches.

(continued)

.North Anna Units 1 and 2 B3321 B 3.3.2-19 eiin.-

Revi si on -G-%.

ESFAS Instrumentation B 3.3.2 BASES ACTIONS C.1, C.2.1, and C.2.2 (continued) experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

The Required Actions are modified by a Note that allows one train to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing, provided the other train is OPERABLE. This allowance is based on the reliability analysis assumption of Reference 8 that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the average time required to perform channel surveillance.

D.1. D.2.1, and D.2.2 Condition D applies to:

e Containment Pressure-High; e Pressurizer Pressure-Low Low; 0 Steam Line Differential Pressure-High;

" High Steam Flow in Two Steam Lines Coincident With Tavg-Low.

Low or Coincident With Steam Line Pressure-Low;

" Containment Pressure-Intermediate High High;

" SG Water level-Low Low; "4

" SG Water level-High High (P-14~~

.~.-

'ZW ~ .0# ~ I one- channel is inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are allowed to

'7 C. tArestore the channel to OPERABLE status or to place it in the tripped condition. Generally this Condition applies to Ib_14*" .61-elA functions that operate on two-out-of -three logic. Therefore, failure of one channel places the Function in a.

two-out-of-two configuration. One channel must be tripped to place the Function in a one-out-of-two configuration that satisfies redundancy requirements.

Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requires the unit be placed in MODE 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(continued)

North Anna Units 1 and 2B332-7Rvso B 3.3.2-37 Revision tti_

ECCS-Operati ng B 3.5.2 BASES SURVEILLANCE SR 3.5.2.4 (continued)

REQUIREMENTS which encompasses the ASME Code. The ASME Code provides the activities and Frequencies necessar'y to satisfy the requirements.

SR 3.5.2.5 and SR 3.5.2.6-These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on-an actual or simulated SI signal and that each ECOS pump capable of starting automatically starts on receipt of an actual or simulated SI signal. 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 these Surveillances under the conditions that apply during a unit outage and the potential for unplanned unit transients if the Surveillances were performed with the reactor at power.

The 18 month Frequency is also acceptable based on.

consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.

SR 3.5.2.7 Proper throttle valve position is necessary for proper ECCS performance and to prevent pump runout and subsequent componqnt damage.; The Surveillance verifies each listed ECOS throttle Valve is secured in the correct position. The 18 month Frequency is based on the same reasons* as those stated in SR 3.5.2.5 and SR 3.5.2.6.

SR .3.5.2.8 4~I3II Periodic inspect* s< of the containment sump suet ~-44et ensure that P:-1ýeunrestricted and stay)'in proper oper'a'ting 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.

North Anna Units 1 and 2 B3521 B 3.5.2-10 Rev i s i on eiin-~

Containment B 3.6.1 BASES BACKGROUND b. Each air lock is OPERABLE, except as provided in (continued) LCO 3.6.2, "Containment Air Locks";

c. All equipment hatches are closed; and

d. The sealing mechanism associated with each penetration (e.g. welds, bellows, or 0-rings) is OPERABLE.

APPLICABLE The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and.temperatures of the limiting DBA without exceeding the design leakage rate.

The. DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a LOCA, a steam line break, and a rod ejection accident (REA)

(Ref. 2). In addition, release of significant fission product radioactivity within-containment can occur from a LOCA or REA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.1% of containment air weight per day (Ref. 3). This leakage rate, used to evaluate offsite doses resulting from accidents, is defined in 1O*-CFR 50, Appendix J, Option B (Ref. 1), as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) resulting from the limiting design basis LOCA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing. La is assumed to be 0.1% of containment air weight per day in the safety'analyses at Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.

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

LCO Containment OPERABILITY is maintained by limiting leakage to

• 1.0 Lag except prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test. At this time the applicable leakage limits must be met.

(cont~inued)

North Anna Units 1 and 2 B3612Rvso B 3.6.1-2 0 Revision -

Containment Air Locks B 3.6.2 BASES APPLICABLE The DBAs that result in a release of radioactive material SAFETY ANALYSES within containment are a loss of coolant accident and a rod ejection accident (Ref. 3). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.1% of containment-air weight per day (Ref. 2). This leakage rate is defined in 10 CFR 50, Appendix J, Option B (Ref . 1), as La = 0.1%.of contai nment ai r wei ght per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure Pa ffollowing a design basis LOCA. This allowable leakage ra n~rms the basis for the acceptance criteria imposed on the SRs associated with the air locks.

The containment air locks satisf~y Criterion 3 of 10 CFR

50. 36 (c)(2)(ii).

LCO Each containment air lock forms part of the containment pressure boundary. As part of the containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. Opening or closing of the manways of the 7 ft personnel air lock is treated in the same manner as opening or closing of the associated door. The interlock allows only one air lock door of an air lock to be opened at one time.

Operation of the manways of the 7 ft personnel air lock is controlled administratively. These provisions ensure that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for entry into or exit from containment.

North Anna Units 1 and 2 B3622Rvso B 3.6.2-2 O Revi si on -9

Contai nment Pressure B 3.6.4 B 3.6 CONTAINMENT'SYSTEMS B 3.6.4 'Containment Pressure BASES BACKGROUND Containment air partial pressure is a process variable that is monitored and controlled. The containment air partial pressure is maintained as a function of refueling water storage tank temperature and service water temperature according to Figure 3.6.4-1 of the LCO, to ensure that, following a Design Basis Accident (OBA),.the containmentt~c ~lV would depressurize i11 < 66 ininutes Lu subdtlIUJIAVp .

conditions. Controlling containment partial pressure within prescribed limits also prevents the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of an inadvertent actuation of the Quench Spray (QS)

System.

The containment internal air partial pressure limits of Figure 3.6.ý4-1 are derived from the input conditions used in the containment DBA analyses. Limiting the containm 'ent internal air partial pressure and temperature in turn limits the pressure that could be expected following a OBA, thus ensuring containment OPERABILITY. Ensuring containment OPERABILITY limits leakage of fission product radioactivity from containment to the environment.

APPLICABLE Containment air partial pressure is an initial condition SAFETY ANALYSES used in the containment DBA analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered relative to containment pressure are the loss of coolant accident (LOCA).and steam line break (SLB).

The LOCA and SLB are analyzed using-computer codes designed to predict the resultant containment pressure transients.

DBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed assuming degraded containment Engineered Safety Feature (ESF) systems (i.e., assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure, resulting in one train of the QS System and one train of the Recirculation Spray System becoming inoperable). The containment analysis for the DBA (Ref. 1) shows that the maximum peak containment pressure results from the limiting design basis SLB.

(conti nued)

North Anna Units 1 and 2B364-ReionO-B 3.6.4-1 Revi si on

Containment Pressure B 3.6.4 BASES (ýFDZIX APPLICABLE The maximum design *nternal pressure for the containment. is SAFETY ANALYSES 45.0 psig. The LOCAlainalyses establish the limits for the (continued) containment air partial pressure operating range. The initial conditions used in the containment design basis LOCA analyses were an air partial pressure of h1-ýpsia a~nd anýair temperature of ,F. This resulted in a maximu peak con ainm~eninternal pressure of .44.-i-psig, which is less than the maximum design internalpesrfo the-containment.ý;

I -T 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 psi and Iz 01F. This resulted i~n a minimum pressure

'1, n-ýide contai ment of.*7(!psia, which is considerably above the design m/ji.um of .5.p a.-

7 For certain aspects of tran-s-ir -accident analyses, maximizing the calculated containment pressure is not

!7 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 the reflood phase calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50, Append ix K (Ref. 2).

The radiological onsequences analysis'demonstrates acceptable resu s provided the containment pressurex-6D .

decreases to . 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 exceed4e--5-psig for the interval from 1 to hours following the Design Basis.<

Accident (Ref. 3). Beyond hours the containment pressure is assumed to be less than 0.0 psig, terminating leakage(

from containment.

Containment pressure satisfies Criterion 2 of 10 CFR

50. 36 (c)(.2) (ii).

LCO Maintaining containment pressure within the limits shown in Figure 3.6.4-1 of the LCO ensures that in the event of a DBA the resultant peak containment, accident pressure will be maintained below the containment design pressure. These limits also prevent the containment pressure from exceeding (conti nued)

North Anna Units 1 and 2B36.-Reionr B 3.6.4-2 Revision k1r7

Containment Pressure B 3.6.4 BASES LCO the containment design negative pressure differential with (continued) respect to the outside atmosphere in the event of inadvertent actuation of the QS System. The LCO limits also APPLICABILITY In MODES 1, 2,. 3, and 4, a DBA could cause a release of-radioactive material 'to containment. Since maintaining containment pressure within design basis-limits is essential to ensure initial conditions assumed in the accident analyses are maintained, the LCO is applicable in MODES 1, 2, 3, and 4.

In MODES 5 and 6, the probability and consequences of these events are-.reduced due to the Reactor Coolant System pressure and temperature limitations of these MODES.

Therefore, maintaining contai~nment pressure within the limits of the LCO is not required in MODE 5 or 6.

ACTIONS A.1 When containment air partial pressure is not within the limits of the LCO, containment pressure must be restored to within these limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment,'"

which requires that containment be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B.1 and B.2 If cont ainment 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.

North Anna Units 1 and 2B3..-Reion2'B 3.6.4-3 Revision

Containment Air Temperature B 3.6.5 B 3.6 CONTAINMENT SYSTEMS.

B 3.6.5 Containment Air Temperature BASES BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). The containment average air temperature is'1limi 'ted during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB).

The containment-average air temperature limit is derived from the input conditions used in the containment functional analyses and the containment structure external pressure analyses. This LCO ensures that initial conditions assumed in the analysis of containment response to a DBA are not violated during unit operations. The total amount of energy to be removed from containment by the Containment Spray .4P4 ee1jsystems during post accident conditions is.dependent upon the energy released to the containment due to the event, as well as the initial containment temperature and pressure.

The higher the initial temperature, the more energy which must be removed, resulting in a higher peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis. Operation with containment temperature in excess of the LCO limit violates an initial condition assumed in the accident analysis.

APPLICABLE Containment average air temperature is an initial condition SAFETY ANALYSES used in the DBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. The limit for containment average air temperature ensures that operation is maintained within the assumptions used in the DBA analyses for containment (Ref. 1).

The limiting DBAs considered relative to containment OPERABILITY are the LOCA and SLB. The DBA LOCA and SLB are analyzed using computer codes designed to preditt the resultant containment pressure transients.-No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed with regard to containment (conti nued)

North Anna Units 1 and-2 B '3.6.5-1 B3651Rvso Revision-@-

Containment Air Temperature B 3.6.5 BASES APPLICABLE Engineered Safety Feature (ESF) systems, assuming no offsite SAFETY ANALYSES power and the loss of one emergency diesel generator, which (continued) is the worst case single active failure, resulting in one train of the' Quench Spray (QS) System and RecirculatiqjL_,..--

Spray System being rendered inoperable. ~ _

The limiting DBA for the maximum peak containment air temperature is an SLB. The initial containment avera eAai~r temperature assumed in the design basis analyses is F.~

This resulted in a maximum containment air temperature of

.5ig a-'ýF.The design temperature is 2800F.

The temperature upper limit is used to establish the environmental qualification operating envelope for containment. The maximum peak containment air temperature was 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 the performance of safety related equipment inside containment (Ref. 2). Thermal analyses showed that the time interval during which the containment air temperature exceeded the containment design temperature was short enough that there would be no adverse effect on equipment inside containment assumed to mitigate the consequences of the DBA.

Therefore, it is concluded that the calculated transient containment air temperature is acceptable for the DBA SLB.

The temperature upper limit is also used in the depressurization analyses to ensure that the minimum pressure limit is maintained following an inadvertent actuation of the QS System (Ref.. 1).

The containment pressure transient is sensitive to the initial air mass in containment and, therefore, to the initial containment air temperature. The limiting DBA for establishing the maximum peak containment internal pressure is an SLB. The- temperature upper limit is used in the SLB analysis to ensure that, in the event of an accident, the maximum containment internal pressure will not be exceeded.

Containment average air temperature satisfies Criterion 2 of 10 CFR 50.36(ic) (2)(ii).

North Anna Units 1 and 2B36.-Reion- B 3. 6. 57 2 Revi si on -%

QS System B 3.6.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Quench Spray (QS) System BASES BACKGROUND The QS System is designed to provide containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values. The QS System, operating in conjunction with the Recirculation Spray (RS)

System, is designed to cool. and depressurize the containment structure to zubatfflasphei ic prez-urcin legg than 6f ig~

Redu~ct~ion -of fol lowing a Design Basis A~ccide'nt -(OBA).'

containment pressure and the iodine removal capability of~

the spray limit the release of fission product radioactivity from containment to the environment in the event of a DBA.

The QS System consists of two separate trains of equal capacity, each capable of meeting the design bases. Each train includes a spray pump, a ded~icated spray header, nozzles., valves, and piping. Each train is powered from a separate Engineered Safety Features (ESF) bus. The refueling water storage tank (RWST) supplies borated water to the QS System.

The QS System is actuated either automatically by a containment High-High pressure signal or manually. The QS System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature during a DBA. Each train of the QS System provides adequate spray coverage to meet the system design requirements for containment heat and iodine fission product removal. The QS System also provides flow to the Inside RS pumps to improve the net positive suction head available.

The Chemical Addition System supplies a sodium hydroxide (NaOH) so'u 'tion into the spray. The resulting alkaline pH of the spray enhances the ability of the spray to scavenge iodine fission products from the containment atmosphere. The NaGH added to the spray also ensures an alkaline pH for the

• solution recirculated in the containment sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and minimizes the occurrence of 'chloride and caustic stress corrosion on mechanical systems and components

• exposed to the fluid.

(conti nued)

North Anna Units 1 and 2 B 3.6.6-1 Revision ,

. 0

QS System B 3.6.6 BASES BACKGROUND The QS System is a containment ESF system. It is designed to (continued) ensure that the heat removal capability required during the post accident period can be attained. Operation of the QS System and RS System provides the required heat removal capability to limit post accident conditions to less than the containment design values and depressurize the containment structure to sub- .I-at-FLI I--ricI rcAzU I

- 6 ii~iue~following a DBA.

The QS.System limits the temperature 'and pressure that could be expected following a OBA and ensures that containment leakage is maintained consistent with the accident analysis.

APPLICABLE The limiting DBAs considered are the loss of coolant AFETY ANALYSES S. accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, with respect to containment ESF Systems, assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure, resulting in one train of the QS System and the RS System inoperable._

During normal operation, the containment internal pressure is varied, along with other parameters, to maintain the capability to depressurize the containment =

ubano~hcrh p~ssuc i 60mintz~after a DBA. This capability and the variation of containment pressure during a DBA are functions of the service water temperature, the RWST water temperature, and thn ,o~tainmentair temperature.

The DBA analyses (Ref. 1)(show' the maximum peak containment pressure of 4 -psig results from the SLB analysis and is calculated to be less than the containment design pressure. The maximum peak containment atmosphere

~-~temperatureoof_*WF results from the SLB analysis and was

-- ctum-e86ece 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 exceeded the containment design temperature was short enough that there would be no adverse effect on equipment inside containment assumed to mitigate the consequences of the DBA.

(conti nued)

North Anna Units I-and 2'B3662Rvso B 3.6.6-2 Revision 0

QS System B 3.6.6 BASES APPLICABLE Therefore, it is concluded that the calculated transient SAFETY ANALYSES containment atmosphere temperatures are acceptable for the (continued) 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 analy~ses of .

peak calculated containment temperature and pressure responses. The QS System total response time of ~i-~secondst comprises the signal delay, diesel generator startup time, and system startup time, including pipe fill time. A For certain aspects of accident analyses, maximizing the 4rcs-r<.-

calculated containment pressure is not conservative. In Ay44 Af4j 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 accordance with 10 CFR 50, Appendix K (Ref. 3).

Inadv ertent 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 contaii -nt minimum pressure.

The radiological (onsequences analysis demonstrates so acceptable resul ij provided the containment pressure .

decreases to &e.-5(lpsig 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 67Xpsig for the interval from 1 to hours following the Design Basis Accident (Ref. 4). Beyond hours the containment pressure I is assumed to be less tha 0o.0 psig, terminating leakage from containment.'

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

LCO During a DBA, one train of the QS Sy stem is required to-provide the heat removal capability assumed in the safety-.

analyses for containment. in addition, one QS System train, with spray pH adjusted by the contents of the chemical addition tank, is required to scavenge iodine fission (continued)

North Anna Units 1 and 2B366-Reion26 B 3.6.6-3 Revision zle>

• RS System B 3.6.7 B 3.6 CONTAINMENT SYSTEMS B 3.6.7 Recirculation Spray (RS) System BASES BACKGROUND The RS System, operating in conjunction with the Quench

.Spray (QS) System, is designed to limit the post accident pressure and temperature in the containment to less than the design values and to depressurize the containment structurp---

to ý--6atf.-sheri --i ess han 8 /iu+--

follo-wi-ng-a-Design Bvasis Accident (DBA). The reduction o Af!U containment pressure and the removal of iodine from the Z&t containment atmosphere by the spray limit the release of fission product radioactivity from containment to the environment in the event of a DBA.

The RS System consists of two separate trains of equal capacity, each *capable of meeting the design and accident analysis bases. Each train includes onie-RS subsystem outside containment and one RS subsystem inside containment. Each subsystem consists of one approximately 50% capacity spray pump, one spray cooler, one 1800 coverage spray header, nozzles, valves, piping, instrumentation, and controls. Each outside'RS subsystem also includes a casing cooling pump with its own valves, piping, instrumentation, and controls.

The two outside RS subsystems' spray pumps are located outside containment and the two inside RS subsystems' spray pumps are located inside containment. Each RS train (one inside and oine outside RS subsystem) is powered from a separate Engineered Safety Features (ESF) bus. Each train of the RS System provides adequate spray coverage to meet the system design requirements for containment heat and iodine fission product removal. Two spray pumps are required to provide. 3600 of. containment spray coverage assumed in the accident analysis. One train of RS or two outside RS subsystems will provide the containment spray coverage and required flow.

The two casing cooling pumps and common casing cooling tank are designed to increase the net positive suction head (NPSH) available to the outside RS pumps by injecting cold water into the suction of the spray pumps. They are also beneficial to the containment depressurization analysis. The casing cooling tank contains at least 116,500 gal of chilled and borated water. Each casing cooling pump supplies one outside spray pump with cold borated water frPm the casing (conti nued)

.North Anna Units 1 and 2 B 3.6.7-1 B3671Rvso Revision-6-?

RS System B 3.6.7 BASES BACKGROUND cooling tank. The casing cooling-pumps are considered part (continued) of the outside RS subsystems. Each casing cooling pump is powered fr-am a separate ESF bus.

The inside RS subsystem pump NPSH is increased by reducing the temperature of the water at the pump suction. Flow is

• diverted from the QS system to the suction -of the inside RS pump on the same safety train as the quench spray pump

• supplying the water.

The RS System provides a spray of subcooled water intoth upper regions of containment to reduce the containment pressure and temperature during a DBA. Upon receipt of a High-High containment pressure signal, the two casing cooling pumps start, the casing cooling discharge valves open, and the RS pump suction and discharge v yes receive an open signal to assure the valves are open. A*re 4-0-0+/-L5 ----- --

imc delay, the insidc RS pump: start, an-aftcr ~ ~

eodtm b .. delay,.'the out-sidc R8 pump- start e1~

The RS pumps take suction from .the containment sump and discharge through their respective spray coolers to the spray headers and into the containment atmosphere. Heat is transferred from the-containment sump water to service water lin'the spray coolers.

The Chemical Addition System supplies a sodifum hydroxide (NaOH) solution to the RWST water supplied to the suction of the QS System pumps. The NaOH added to the QS System spray ensures an alkaline pH for the solution recirculated in the containment sump. The resulting alkaline pH of the RS spray

• (pumped from the sump) enhances the ability of the spray to scavenge iodine fission products from the containment atmosphere. The alkaline pH of the containment sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

The RS System is a containment ESF system. It is designed to ensure that the heat removal capability required during the post accident period can be attained.'Operation of the QS and RS systems provides the required heat removal. capability to design values and depressurize the containment structure t-e' The RS System limits the temperature and pressure that could be expected following a OBA and ensures that containment leakage is maintained consistent with the accident analysis..

North Anna Units 1 and 2 B3672AedetNs B 3.6-7-2 ~2A-Amendment Nos. 2ý.

RS System B 3.6.7 BASES APPLICABLE The limiting DBAs considered are the 4ns of cool ait SAFETY ANALYSES .acz4-d~.t 4LOCAY and the steam line brca-k A'LB7'. The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients; DBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure for containment depressurization, resulting in one train of the~-----x---

QS and RS systems being rendered inoperable (Ref.1)

The peak containment pressure following a high energy line break is affected by the initial total pressure and temperature of the containment atmosphere and the QS System operation. Maximizing the initial containment total pressure and average atmospheric temperature maximizes the calculated peak pressure. The heat removal effectiveness of the QS System spray is dependent on the temperature of the water in the refueling w;ater storage-tank d(RWS1J.y. The time required to depressurize the containment and thie capability to maintain it depressurized below atmospheric pressure depend on the functional performance of the QS and RS systems and the service water temperature. When the Service Water temperature is elevated, it is more difficult to depressurize the containment within 60 mntssince th e Z-. RIa*

rid heat removal effectiveness of the RS System -- is limited. -~ie~

During normal operation, the containment internal pressure is varied to maintain the capability to depressurize the containment tta a~b~uki rzuei cst~

69O-i~i-tes after a DBA. This capability and the variation of containment pressure are functions of service water temperature, RWST water temperature, and the containment air temperature.

Theanayses DA t the maximum peak containment tow pressure of 2 psig results from the SLB analysis and is calculated to be less than the containment design pressure.

Th aitm751 peak containment atmosphere temperature results from the SLB analysis and is calculated to exceed the

=309 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 (conti nued)

North Anna Units 1 and 2 B3673Rvso B 3.6.7-3 Revision 0

RS System B 3.6.7 BASES APPLICABLE the containment atmosphere temperature exceeds the SAFETY ANALYSES containment design temperature is short enough that there (continued) would be no adve*rse 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 rom the containment sis(

is based upon a nse time associated wit eeding the(

High-Hi ainment pressure signal .oint to achieving ow through the RS Syste ay nozzles. A delay in response time initiatio vides conservative analys peak calculated inment temperature and pr e. The RS System's I respon se time is determ* y the delay 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 accordance with 10 CFR 50, Appendix K (Ref. 3).

The radiological consequences analysis demonstrates.

~ acceptable results provided the containment pressure

ýý ~d~ecreas~es to .9 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 6_5 psig trthe inte~rval from 1 to hours following the Design Basis%

Accident (Ref. 4). Beyond~ hours the containment pressure is assumed to be less thjan[0.0 psig, terminating leakage from containmen~t.

The RS System satisfies Oriterion 3 of 10 CFR

50. 36(c) (2)(ii).

LCO During a DBA, one train (one inside and one outside RS subsystem in the same train) or two outside RS subsystems of the. RS System are required to provide the minimum heat removal capability assumed in the safety analysis. To ensure-that this requirement is met, four RS subsystems and the casing cooling tank must be OPERABLE. This will ensure that at least one train will'operate assuming the worst case single failure occurs, which is no offsite power and the loss of one emergency diesel generator. Inoperability of the (continued)

North Anna Units 1 and 2B367-Reion.O B 3.6.7-4 Revision .24-

RS System B 3.6.7 BASES SURVEILLANCE SR 3.6.7.4 (continued)

REQUIREMENTS it involves verification, through a system walkdown, that those valves outside containment and capable of potentially being mispositioned are in the correct position.

SR 3.6.7.5 Verifying that each RS and casing cooling pump's developed head at the flow test point is greater than or equal to the required developed head ensures that these pumps' performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 5). Since the RS System pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.

SR 3.6.7.6 These SRs ensure that each automatic valve actuates and that the Rt Syst-tem rbrcasing cooling pumps start upon receipt of .Lzi I,."

an actual or simulated High-High containment pressure ~ r0 signal. t aytmsaa as eiIIAV

-mpuffThis 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 13.from

ý-ýJsaaT a reliability standpoint.

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 (continued)

North Anna Units 1 and 2 B3678Rvso B 3.6-7-8 Revision E&Pý

RS System B 3.6.7 BASES SURVEILLANCE SR 3.6.7 continued)

REQUIREMENTS 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 CFR 50, Appendix K.

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 3.6.7-9 B3679Rvso Revision et>

Serial No.06-849 Docket Nos. 50-338/339 ATTACHMENT 5 PROPOSED TECHNICAL SPECIFICATION CHANGE AND SUPPORTING SAFETY ANALYSES REVISIONS TO ADDRESS GENERIC SAFETY ISSUE 191 TYPED TECHNICAL SPECIFICATION BASES PAGES VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 1. Safety Injection (continued)

SAFETY ANALYSES, LCO, f. g. Safety Injection-High Steam Flow in Two Steam Lines AND Coincident With Tav -Low Low or Coincident With Steam-APPLICABILITY Line Pre ssu re-Low 'econtinued)

With the transmitters located inside the containment (Tavg) or near the steam lines (High Steam Flow), it is possible for them to experience adverse steady state environmental conditions during an SLB event.

The trip setpoint reflects only steady state instrument uncertainties.

This Function must be.OPERABLE in MODES 1, 2, and 3

• (above P-12) when a secondary side break or stuck open valve could result in the rapid depressurization of the steam line(s). This signal may be manually blocked by the operator when below the P-12 setpoint.

Above P-12, this Function is automatically unblocked.

This Function is not required OPERABLE below P-12 because the reactor is not critical, so steam line break is not a concern. SLB may be addressed by Containment Pressure High (inside containment) or by High Steam Flow in Two Steam Lines coincident with Steam Line Pressure-Low, for Steam Line Isolation, followed by High Differential Pressure'Between Two Steam Lines, for SI. This Function is not required to be OPERABLE in MODE 4, 5, or 6 because there is insufficient energy in the secondary side of the unit to cause an accident.

2. Containment Spray Systems The Containment Spray Systems (Quench Spray (QS) and Recirculation Spray (RS)) provide four primary functions:

1. Lowers containment pressure and temperature after an HELB in containment;

~.Reduces the amount of radioactive iodine in the containment atmosphere;

3. Adjusts the pH of the water in the containment sump after a large break LOCA; and

4. Remove heat from containment.I North Anna Units 1 and 2 B3321 B 3.3.2-13

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 2. Containment Spray Systems (continued)

SAFETY ANALYSES, LCO, These functions are necessary to:

AND APPLICABILITY -

• Ensure the pressure boundary integrity of the containment structure;

• Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure; and 9 Minimize corrosion of the components and systems inside containment following a LOCA.

The containment spray actuation signal starts the QS pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment. Water is initially drawn from the RWST by the QS pumps and mixed with a sodium hydroxide solution from the chemical addition tank. When the RWST level reaches the low setpoint coincident with Containment Pressure-High High, the RS pumps receive a start signal.

The outside RS pumps start irmmediately and the inside RS pumps start after a 120-second delay. Water is drawn from the containment sump through heat exchangers and discharged to the RS nozzle headers. When the RWST reaches the low low level setpoint, the Low Head Safety Injection pump suctions are shifted to the containment sump. Containment spray is actuated manually or by Containment Pressure-High High signal. RS is actuated manually or by RWST Level-Low coincident with Containment Pressure-High High.

a. Containment Spray-Manual Initiation The operator can initiate containment spray at any time from the control room by simultaneously turning two containment spray actuation switches in the same train. Because an inadvertent actuation of containment spray could have such serious consequences, two switches must be turned simultaneously to initiate containment spray. There are two sets of two switches each in the control room.

(continued)

North Anna Units 1 and 2 B3321 B 3.3.2-14

ESFAS Instrumentation B 3.3.2 BASES APPLI CABLE 2. Containment Spray Systems (continued) I SAFETY ANALYSES, LCO, a. Containment Spray-Manual Initiation (continued)

AND APPLICABILITY Simultaneously turning the two switches in either set will actuate containment spray in both trains~in the same manner as the automatic actuation signal. Two Manual Initiation switches in each train are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function. Note that Manual Initiation of containment spray also actuates Phase B containment isolation.

b. Containment Spray-Automatic Actuation Logic and Actuation Relays Automatic actuation logic and actuation relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b.

Manual and automatic initiation of containment spray must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and sufficient energy exists in the primary or secondary systems to pose a threat to containment integrity due to overpressure conditions. Manual initiation is also required in MODE 4, even though automatic actuation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a OBA. However, because of the large number of components actuated on a containment spray, actuation is simplified by the use of the manual actuation switches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system manual initiation. In MODES 5 and 6, there is insufficient energy in the primary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond, to mitigate the consequences of abnormal conditions by manually starting individual components.

North Anna Units 1 and 2 B3321 B 3.3.2-15

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 2. Containment Spray Systems (continued) I SAFETY ANALYSES, LCD, c. Containment Spray-ContainmentPressure AND APPLICABILITY This signal provides protection against a LOCA or an SLB inside containment. The transmitters (d/p cells) are located outside of containment with the sensing line (high pressure side of the transmitter) located-inside containment. The transmitters and electronics are located outside of containment. Thus, they will not experience any adverse 'environmental conditions and the Allowable Value reflects only steady state instrument uncertainties.

This is one of few Functions that requires the bistable output to energize to perform its required action. It is not desirable to have a loss of power actuate containment spray, since the consequences of an inadvertent actuation of containment spray could be serious. Note that this Function also has the inoperable channel placed in bypass rather than trip to decrease the probability of an inadvertent actuation.

North Anna uses four channels in a two-out-of-four logic configuration and the Containment Pressure-High High Setpoint Actuates Containment Spray Systems.

Since containment pressure is not used for control, this arrangement exceeds the minimum redundancy requirements. Additional redundancy is warranted because this Function is energize to trip.

Containment Pressure-High High must be OPERABLE in MODES 1,.2, and 3 when there is sufficient energy in the primary and secondary sides to pressurize the containment following a pipe break. In MODES 4, 5, and 6, there is insufficient energy in the primary and secondary sides to pressurize the containment and reach the Containment Pressure-High High setpoints.

d. RWST Level-Low Coincident with Containment Pressure-High High This signal starts the RS system to provide protection against a LOCA inside containment. The Containment Pressure-High High (ESFASFunction 2.c) signal aligns the RS system for spray flow delivery (e.g., opens isolation valves) but does not start the*

(continued)

North Anna Units 1 and 2 B3321 B 3.3.2-16

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE 2. Containment Spray Systems (continued)

SAFETY>_

ANALYSES, LCQ, d. RWST Level-Low Coincident with Containment AND Pressure-High High (continued)

APPLICABILITY RS pumps. The RWST Level-Low coincident with Containment Pressure-High High provides the automatic start signal for the inside RS and outside RS pumps.

Once the coincidence trip is satisfied, the outside RS pumps start immediately and the inside RS pumps start after a 120-second delay. The delay time is sufficient to avoid simultaneous starting of the RS pumps on the same emergency diesel generator. This ESFAS function ensures that adequate water inventory is present in the containment sump to meet the RS sump strainer functional requirements following a LOCA.

The RS system is not required for SLB mitigation.

Automatic initiation of RS must be OPERABLE in MODES 1, 2, and 3 when there is a potential for an accident to occur, and sufficient energy exists in the primary and secondary systems to pose a threat to containment integrity due to overpressure conditions. The requirement for automatic initiation of RWST Level-Low to be operable in MODES 1, 2, and 3 is consistent with the operability requirements for Containment Pressure-High High. Manual initiation of the RS system is required in MODE 4, even though automatic initiation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA. In MODES 5 and 6, there is insufficient energy in the primnary and secondary systems to result in containment overpressure. In MODES 5 and 6, there is also adequate time for the operators to evaluate unit conditions and respond to mitigate the consequences of abnormal conditions by manually starting individual components. An operator can initiate RS at any time from the control room by using the pump control switch. The manual function would be used only when adequate water inventory is present in the containment sump to meet the RS sump strainer functi onal requirements.

North Anna Units 1 and 2 B3321 B 3.3.2-17

ESFAS Instrumentation B 3.3.2 BASES APPLI CABLE 3. Containment Isolation (continued)

SAFETY ANALYSES, LCO, a. Containment Isolation-Phase A Isolation (continued)

AND APPLICABILITY Manual and automatic initiation of Phase A Containment Isolatio~n must be OPERABLE inMODES 1, 2, and 3, when there is a potential for an accident to occur. Manual initiation is also required in MODE 4 even though automatic actuation is not required. In this MODE, adequate time is available to manually actuate required components in the event of a DBA, but because of the large number of components actuated on a Phase A Containment Isolation, actuation is simplified by the use of the manual actuation switches. Automatic actuation logic and actuation relays must be OPERABLE in MODE 4 to support system manual initiation. In MODES 5 and 6, there is insufficient energy in the primary or secondary systems to pressurize the containment to require Phase A Containment Isolation. There also is adequate time for the operator to evaluate unit conditions and manually actuate individual isolation valves in response to abnormal or accident conditions.

(3)Phase A Isolation-Safety Injection Phase A Containment Isolation is also initiated by all Functions that initiate SI. The Phase A Containment Isolation requirements for these Functions are the same as the requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

b. Containment Isolation-Phase B Isolation Phase B Containment Isolation is accomplished by Manual Initiation, Automatic Actuation Logic and Actuation Relays, and by Containment Pressure channels (the same channels that actuate Containment Spray Systems, Function 2). The Containment Pressure I trip of Phase B Containment Isolation is energized to trip in order to minimize the potential of spurious trips that may damage the RCPs.

(1)Phase B Isolation-Manual Initiation North Anna Units 1 and 2 B3322 B 3.3.2-20

ESFAS Instrumentation B 3.3.2 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)

" High Steam Flow in Two Steam Lines Coincident With Tavg-Low Low or Coincident With Steam Line Pressure-Low;

• Containment Pressure-Intermediate High High;

" SG Water Level-Low Low;

" SG Water Level-High High (P-14); and

" RWST Level-Low Coincident With Containment Pressure High High.

If one channel is inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are allowed to restore the channel to OPERABLE status or to place it in the tripped condition. Generally this Condition applies to functions that operate on two-out-of-three logic. Therefore, failure of one channel places the Function in a two-out-of-two configuration. One channel must be tripped to place the Function in a one-out-of-two configuration that satisfies redundancy requirements.

Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requires the unit be placed in MODE 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 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. In MODE 4, these Functions are no longer required OPERABLE.

The Required Actions are modified by a Note that allows the inoperable channel to be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed to restore the channel to OPERABLE status or to place the inoperable channel in the tripped condition, and the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for testing, are justified in Reference 8.

E.1, E.2.1, and E.2.2 Condition E applies to:

• Containment Spray Containment Pressure-High High; and North Anna Units 1 and 2 B 3.3.2-38 B3323

EGOS-Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.4 (continued)

REQU IREMENTS which encompasses the ASME Code.. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and- that each ECCS pump capable of starting automatically starts on receipt of an actual or simulated SI signal. 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 these Surveillances under the conditions that apply during a unit outage and the potential for unplanned unit transients if the Surveillances were performed with the reactor at-*power.

The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.

SR 3.5.2.7 Proper throttle valve position is necessary for proper EGGS performance and to prevent pump runout and subsequent component damage. The Surveillance verifies each listed EGGS throttle valve is secured in the correct position. The 18 month Frequency is based on the same reasons as those stated in SR 3.5.2.5 and SR 3.5.2.6.

SR 3.5.2.8 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.

North Anna Units 1 and 2 B3521 B 3.5.2-10

Contai nment B 3.6.1 BASES BACKGROUND b. Each air lock is OPERABLE, except as provided in (continued) LCO 3.6.2, "Containment Air Locks";

c. All equipment hatches are closed; and

d. The sealing mechanism associated with each penetration (e.g. welds, bellows, or 0-rings) is OPERABLE.

APPLICABLE The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a LOCA, a steam line break, and a rod ejection accident (REA)

(Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA or REA. In the OBA*analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.1% of containment air weight per day (Ref. 3). This leakage rate, used to evaluate offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J, Option B (Ref.- 1), as La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) resulting from the limiting design basis LOCA. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing. La is assumed to be 0.1% of containment air weight per day in the safety analyses at Pa (Ref. 3).. I Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.

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

LCO Containment OPERABILITY ismaintained by limiting leakage to

• 5 1.0 La, except prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test. At this time the applicable leakage limits must be met.

(continued)

North Anna Units 1 and 2B361- B 3.6.1-2

Containment Air Locks B 3.6.2 BASES APPLICABLE The DBAs that result in a release of radioactive material SAFETY ANALYSES within containment are a loss of coolant accident and a rod ejection accident (Ref. 3). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.1% of containment air weight per day (Ref. 2). This leakage rate is defined in 10 CFR 50, Appendix J, Option B (Ref. 1), as La = 0.1% of contai nment ai r wei ght per day, the maximum allowable containment leakage rate at the calculated peak containment internal pressure Pa following a design I basis LOCA. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.

The containment air locks satisfy Criterion 3 of 10 CFR 50.36(c) (2)(ii).'

LCO Each containment air lock forms part of the containment pressure boundary. As part of the containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

Each ai r lock is requi red to be OPERABLE. For the ai r lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the ai r lock must be in compl iance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. Opening or closing of the manways of the 7 ft personnel air lock is treated in the same manner as opening or closing of the associated door. The interlock allows only one air lock door of an air lock to be opened at one time.

Operation of the manways of the 7 ft personnel air lock is controlled administratively. These provisions ensure that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single door in each air lock-is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for entry into or exit from containment.

North Anna Units 1 and 2B362- B 3.6.2-2

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND Containment air partial pressure is a process variable that is monitored and controlled. The containment air partial pres.sure is maintained as a function of refueling water storage tank temperature and service water temperature according to Figure 3.6.4-1 of the LCO, to ensure that, following a Design Basis Accident (DBA), the containment would depressurize to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Controlling I containment partial pressure within prescribed limits also prevents the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of an inadvertent actuation of the Quench Spray (QS) System.

Controlling containment air partial pressure limits within prescribed limits ensures adequate net positive suction head (NPSH-) for the recirculation spray and low head safety injection pumps following a DBA.

The containment internal air partial pressure limits of Figure 3.6.4-1 are derived from the input conditions used in the containment DBA analyses. Limiting the containment internal air partial pressure and temperature in turn limits the pressure that 'could be expected following a DBA, thus ensuring containment OPERABILITY. Ensuring containment OPERABILITY limits leakage of fission product radioactivity from containment to the environment.

APPLICABLE Containment air partial pressure is an initial condition SAFETY ANALYSES used in the containment OBA analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered relative to containment pressure are the loss of coolant accident (LOCA) and steam line break (SLB).

The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure transients.

DBAs are assumed not to occur simultaneously or.

consecutively. The postulated DBAs are analyzed assuming degraded containment Engineered Safety Feature (ESF) systems (i.e., assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure, resulting in one train of the QS System and (continued)

North*Anna Units 1 and 2B364- B 3.6.4-1

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 phase of a LOCA analysis increases with increasing containment backpressure. For the reflood phase calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 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 2B3.4- B 3.6.4-2

Containment Pressure B 3.6.4 BASES APPLICABLE 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 SAFETY ANALYSES Accident (Ref. 3). Beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the containment pressure I (continued) is assumed to be less than 0.0 psig, terminating leakage from containment..

Containment pressure satisfies Criterion 2 of 10 CFR 50.36 (c)(2)0ii) .

LCO Maintaining containment pressure within the limits shown in Figure 3.6.4-1 of the LCO ensures that in the event of a DBA the resultant peak containment accident pressure will be maintained below the containment design pressure. These limits also prevent the containment pressure from exceeding the containment design negative-pressure differential with respect to the outside atmosphere in the event of inadvertent actuation of the QS System. The LCO limits also ensure the containment structure will depressurize to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a DBA.

APPLICABILITY In MODES .1,2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within design basis limits is essential to ensure initial conditions assumed in the accident analyses are maintained, th~e LCO is applicable in MODES 1, 2, 3, and 4.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the Reactor Coolant System pressure and temperature limitations of these MODES.

Therefore, maintaining containment pressure within the limits of the LCO is not required in MODE 5 or 6.

ACTIONS A.1 When containment air partial pressure is not within the limits of the LCO, containment pressure must be restored to within these limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment,"

which requires that containment be restored to OPERABLE' status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

North Anna Units 1 and 2B3. 4-B 3.6.4-3

Containment Air Temperature B 3.6.5 B 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperature BASES.

BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss *of coolant accident (LOCA) or steam line break (SLB).

The containment average air temperature limit is derived from the input conditions used in the containment functional analyses and the containment structure external pressure analyses. This LCO ensures that initial conditions assumed in the analysis of containment response to a DBA are not violated during unit operations. The total amount of energy to be removed from containment by the Containment Spray systems during post accident conditions is dependent upon the energy released to the containment due to the event, as well as the initial containment temperature and pressure.

The higher the initial temperature, the more energy which must be removed, resulting in a higher peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis. Operation with containment temperature in excess of the LCO limit violates an initial condition assumed in the accident analysis.

APPLICABLE Containment average air temperature is an initial condition SAFETY ANALYSES used in the DBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. The limit for containment average air temperature ensures that operation is maintained within the assumptions used in the DBA analyses for containment (Ref. 1).

The limiting DBAs considered relative to containment OPERABILITY are the LOCA and SLB. The DBA LOCA and SLB are analyzed-using computer codes designed to predict the resultant containment pressure transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed with regard to containment (conti nued)

North Anna Units 1 and 2B3651 B 3.6.5-1

Containment Air Temperature B 3.6.5 BASES APPLICABLE Engineered Safety Feature (ESF) systems, assuming no off~site SAFETY ANALYSES power and the loss of one emergency diesel generator, which (continued) is the worst case single active failure, resulting in one train of the Quench Spray (QS) System and Recirculation Spray System being rendered inoperable. The postulated SLB events are analyzed without credit for the RS system.

The limiting DBA for the maximum peak containment air temperature is an SLB. The initial containment average air temperature assumed in the design basis analyses is 1151F.

This resulted in a maximum containment air temperature of 309 0F. The design temfperature is 280 0F.

The temperature upper limit is used to establish the environmental qualification operating envelope for containment. The maximum peak containment air temperature was 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 the performance -of safety related equipment inside containment (Ref. 2). Thermal analyses showed that the time interval during which the containment air temperature exceeded the containment design temperature was short enough that there would be no adverse effect on equipment inside containment assumed to mitigate the consequences of the DBA.

Therefore, it is concluded-that the calculated transient containment air temperature is acceptable for the DBA SLB.

The temperature upper limit is also used in the depressurization analyses to ensure that the minimum pressure limit is maintained following an inadvertent actuation of the QS System (Ref. 1).

The containment pressure transient is sensitive to the initial air mass in containment and, therefore, to the initial containment air temperature. The limiting DBA for establishing the maximum peak containment internal pressure is an SLB. The temperature upper limit is used in the SLB analysis to ensure that, in the event of an accident, the maximum containment internal pressure will not be exceeded.

Containment average air temperature satisfies Criterion 2 of 10 CFR 50.36(c) (2)(ii).

North Anna Units 1 and 2B36.- B 3.6.5-2

QS System B 3.6.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Quench Spray (QS) System BASES BACKGROUND The QS System is designed to provide containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values. The QS System, operating in conjunction with the Recirculation Spray (RS)

System, is designed to cool and depressurize the containment structure to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a Design I Basis Accident (DBA). Reduction of containment pressure and the iodine removal capability of the spray limit the release of fission product radioactivity from containment to the environment in the event of a DBA.

The QS System consists of two separate trains of equal capacity, each capable of meeting the design bases. Each train includes a spray pump, a dedicated spray header, nozzles, valves, and piping. Each train is powered from a separate Engineered Safety.Features (ESF) bus. The refueling water storage tank (RWST) supplies borated water to the QS System.

The QS System is actuated either automatically by a containment High-High pressure-signal or manually. The QS System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature during a DBA. Each train of the QS System provides adequate spray coverage to meet the system design requirements for containment heat and iodine fission product removal. The QS System also provides flow to the Inside RS pumps to improve the net positive suction head available.

The Chemical Addition System supplies a sodium hydroxide (NaOH) solution into the spray. The resulting alkaline pH of the spray enhances the ability of the spray to scavenge iodine fission products from the containment atmosphere. The NaOH added to the spray also ensures an alkaline pH for the solut~ion recircul'ated in the containment sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

(continued)

North Anna Units 1 and 2B366- B 3.6.6-1

QS System B 3.6.6 BASES BACKGROUND The QS System is a containment ESF system. -Itis designed to (continued) ensure that the heat removal capability required during the post accident period can be attained. Operation of the QS System and RS System provides the required heat removal capability to limit post accident conditions to less than the containment design values and depressurize the containment structure to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a DBA. I The QS System limits the temperature and pressure that could be expected following a DBA and ensures that containment leakage is maintained consistent with the accident analysis.

APPLICABLE The.limiting DBAs considered are the loss of coolant SAFETY ANALYSES accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, with respect to containment ESF Systems, assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure, resulting in one train of the QS System and the RS System inoperable. The postulated SLB events are analyzed without credit for the RS system.

During normal operation, the containment internal pressure is varied, along with other parameters, to maintain the capability to depressurize the containment to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a DBA. This capability and the variation of containment pressure during a DBA are functions of the service water temperature, the RWST water temperature, and' the containment air temperature.

• The DBA analyses (Ref. 1) 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 peak containment atmosphere temperature of 3097F results from the SLB analysis and was calculated to exceed the containment design temperature for I 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 (conti nued)

North Anna Units 1 and 2B366- B 3.6.6-2

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 OBA.

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 bas'ed 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 mna manner designed to conservatively minimize, rather than maximize, the calculated transient containment.

pressures in accordance with 10 CFR 50, Appendix K (Ref. 3).

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)0ii) .

North Anna Units 1 and 2B366- B 3.6.6-3

RS System B 3.6.7 B 3.6 CONTAINMENT SYSTEMS B 3.6.7 Recirculation Spray (RS) System BASES BACKGROUND The RS System, operating in conjunction with the Quench Spray (QS) System, is designed to limit the post accident pressure and temperature in the containment to less than the design values and to depressurize the containment structure to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a Design Basis Accident (DBA). The reduction of containment pressure and the removal of iodine from the containment atmosphere by the spray limit the release of fission product radioactivity from containment to the environment in the event of a DBA.

The RS System consists of two separate trains of equal capacity, each capable of meeting the design and accident analysis bases. Each train includes one RS subsystem outside containment and one RS subsystem inside containment. Each subsystem consists of one approximately 50% capacity spray pump, one spray cooler, one 1800 coverage spray header, nozzles, valves, piping, instrumentation, and controls. Each outside RS subsystem also includes a casing cooling pump with its own valves, piping, instrumentation, and controls.

The two outside RS subsystems' spray pumps are located outside containment and the two inside RS subsystems' spray pumps are located inside containment. Each RS train (one inside and one outside RS subsystem) is powered from a separate Engineered Safety Features (ESF) bus. Each train of the RS System provides adequate spray coverage to meet the system design requirements for containment heat and iodine fission product removal. Two spray pumps are required to provide 3600 of containment spray coverage assumed in the accident analysis. One train of RS or two outside RS subsystems will provide the containment spray coverage and required flow.

The two casing cooling pumps and commnon casing cooling tank are designed to increase the net positive suction head (NPSH) available to the outside RS pumps by injecting cold w*ater into the suction of the spray pumps. Th.ey are also beneficial to the containment depressurization analysis. The casing cooling tank contains at least 116,500 gal of chilled and borated water. Each casing cooling pump supplies one outside spray pump with cold borated water from the casing (continued)

North Anna Units 1 and 2B367- B 3.6.7-1

RS System B'3.6.7 BASES BACKGROUND cooling tank. The casing cooling pumps are considered part (continued) of the outside RS subsystems-. Each casing cooling pump is powered from a separate ESF bus.

The inside RS subsystem pump NPSH is increased by reducing the temperature of the water at the pump suction. Flow is diverted from the QS system to the suction of the inside RS pump on the same safety train as the quench, spray pump supplying the water.

The RS System provides a spray of subcooled water into the upper regions-of containment to reduce the containment pressure and temperature during a DBA. Upon receipt of a High-High containment pressure signal, the two casing cooling pumps start, the casing cooling discharge valves open, and the RS pump suction and discharge valves receive an open signal to assure the valves are open. Refueling water storage tank (RWST) Level-Low coincident with Containment Pressure-High High provides the automatic start signal for the inside RS and outside RS pumps. Once-the coincidence logic is satisfied, the outside RS pumps start immediately

• and the inside RS pumps start after a 120-second delay. The delay time is sufficient to avoid simultaneous starting of the RS pumps on the same emergency diesel generator. The coincident trip ensures that adequate water inventory is present in the containment sump to meet the RS sump strainer functional requirements following a loss of coolant accident (LOCA). The RS system is not required for steam line break (SLB) mitigation. The RS pumps take suction from the containment sump and discharge through their respective spray coolers to the spray headers and into the containment atmosphere. Heat is transferred from the containment sump water to service water in the spray coolers.

The Chemical Addition System supplies a sodium hydroxide (NaOH) solution to the RWST water supplied to the suction of the QS System pumps. The NaOH added to the QS System spray ensures an alkaline pH for the solution recirculated in the containment sump. The resulting alkaline pH of the R5 spray (pumped from the sump) enhances the ability of the spray to scavenge iodine fission products from the containment atmosphere. The alkaline pH of the containment sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

(continued)

North Anna Units 1 and 2B367- B 3.6.7-2

RS System B 3.6.7 BASES BACKGROUND The RS System is a containment ESF system. It is designed to (continued) ensure that the heat removal capability required during the post accident period can be attained. Operation of the QS and RS systems provides the required heat removal capability to limit post accident conditions to less than the containment design values and depressurize the containment structure to less'than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a DBA.

The RS System limits the temperature and pressure that could be expected following a DBA and ensures that containment leakage is maintained consistent with the accident analysis.

APPLICABLE The limiting DBAs considered are the LOCA and the SLB. The I SAFETY ANALYSES LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients; OBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed assuming no offsite power and the loss of one emergency diesel generator, which is the worst case single active failure for containment depressurization, resulting in one train of the QS and RS systems being rendered inoperable (Ref. 1). The postulated SLB events are analyzed without credit for the RS system.

The peak containment pressure following a high energy line break is affected by the initial total pressure and temperature of the containment atmosphere and the QS System operation. Maximizing the initial containment total pressure and average atmospheric temperature maximizes the calculated peak pressure. The heat removal effectiveness of the QS System spray is dependent on the temperature of the water in the RWST. The time required to depressurize the containment I and the capability to maintain it depressurized below atmospheric pressure depend on the functional performance of the QS and RS systems and the service water temperature. When the Service Water temperature is elevated., itlis more difficult to depressurize the containment to less than 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 to subatmospheric pressure within 6 hfours since the heat removal effectiveness of the RS System is limited.

During normal operation, the containment internal pressure is varied to maintain the capability to depressurize the containment to less than 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 to subatmospheric pressure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a DBA. This I (conti nued)

North Anna Units 1 and 2B367- B 3.6.7-3

RS System B 3.6.7 BASES APPLICABLE capability and the v~ariation 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 309OF 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 particul~ar, 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 accordance with 10 CFR 50, Appendix K (Ref. 3).

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 (continued)

North Anna Units 1 and 2B3674 B 3.6.7-4

RS System B 3.6.7 BASES APPLICABLE 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 I SAFETY ANALYSES is assumed to be less than 0.0 psig, terminating leakage (continued) from containment.

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

LCO During a DBA, one train (one inside and one outside RS subsystem in the same train) or two outside RS subsystems of the RS System are required to provide the minimum heat removal capability assumed in the-safety analysis. To ensure that this requirement is met, four RS subsystems and the casi'ng cooling tank must be OPERABLE. This will ensure that at least one train will operate assuming-the worst case single failure occurs, which is no offsite power and the loss of one emergency diesel generator. Inoperability of the casing cooling tank, the casing cooling pumps, the casing cooling valves, piping, instrumentation, or controls, or of the QS System requires an assessment of the effect on RS subsystem OPERABILITY.

Each RS train consists of one RS subsystem outside containment and one RS subsystem inside containment. Each RS subsystem includes one spray pump, one spray cooler, one 1800 coverage spray header, nozzles, valves, piping, instrumentation, and controls to ensure an OPERABLE flow path capable of taking suction from the containment sump.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the RS System.

.I In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the RS System is not required to be OPERABLE in MODE 5 or 6.

ACTIONS A.1 With one of the RS subsystems inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.

The components in this degraded condition are capable of providing at least 100% of the heat removal needs (i.e.*,

approximately 150% when one RS subsystem is inoperable)

(conti nued)

North Anna Units 1 and 2B367- B 3.6.7-5

RS System B 3.6.7 BASES SURVEILLANCE SR 3.6.7.6 REQUIREMENTS (conti nued) 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 CFR 50, Appendix K.

North Anna Units 1 and 2B367- B 3.6.7-9

RS System B 3.6.7 BASES REFERENCES 4. UFSAR, Section 15.4.1.7.

(continued)

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

North Anna Units 1 and 2 B 3.6.7-10 B3671