ML070720043
| ML070720043 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/13/2007 |
| From: | Siva Lingam NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| Lingam, Siva NRR/DORL 415-1564 | |
| Shared Package | |
| ML070740367 | List: |
| References | |
| GSI-191, TAC MD3197, TAC MD3198 | |
| Download: ML070720043 (30) | |
Text
March 13, 2007 Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGES PER GENERIC SAFETY ISSUE (GSI) 191 (TAC NOS. MD3197 AND MD3198)
Dear Mr. Christian:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 250 and 230 to Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station, Unit Nos. 1 and 2. The amendments change the Technical Specifications (TSs) in response to your application dated October 3, 2006, as supplemented by letter dated January 24, 2007.
These amendments revise the TSs and licensing basis to support the resolution of NRCs GSI 191, assessment of debris accumulation on containment sump performance and its impact on emergency recirculation during an accident, and NRC Generic Letter 2004-02.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Siva P. Lingam, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339
Enclosures:
- 1. Amendment No. 250 to NPF-4
- 2. Amendment No. 230 to NPF-7
- 3. Safety Evaluation cc w/encls: See next page
March 13, 2007 Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGES PER GENERIC SAFETY ISSUE (GSI) 191 (TAC NOS. MD3197 AND MD3198)
Dear Mr. Christian:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 250 and 230 to Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station, Unit Nos. 1 and 2. The amendments change the Technical Specifications (TSs) in response to your application dated October 3, 2006, as supplemented by letter dated January 24, 2007.
These amendments revise the TSs and licensing basis to support the resolution of NRCs GSI 191, assessment of debris accumulation on containment sump performance and its impact on emergency recirculation during an accident, and NRC Generic Letter 2004-02.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Siva P. Lingam, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339
Enclosures:
- 1. Amendment No. 250 to NPF-4
- 2. Amendment No. 230 to NPF-7
- 3. Safety Evaluation cc w/encls: See next page DISTRIBUTION: Public LPL2-1 R/F RidsOgcRp RidsNrrDorlLpl2-1(EMarinos)
RidsNrrDeEeeb(GWilson)
RidsNrrDssSsib(MScott)
RidsNrrPMSLingam(hard copy)
RidsNrrDirsItsb(TKobetz)
RidsNrrDssScvb(RDennig)
RidsNrrLAMOBrien(hard copy)
RidsNrrDorlDpr(BSingal)
RidsNrrDssScvb(ASallman)
RidsNrrDraAadb(ABoatright)
RidsAcrsAcnwMailCenter RidsNrrDraAadb(MKotzalas)
RidsRgn2MailCenter(EGuthrie)
GHill (4 hard copies)
RidsNrrDssSsib(DCullison)
Package No.: ML070740367 Amendment No.: ML070720043 Tech Spec No.: ML070740673
- transmitted by memo dated.
OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/SCVB/BC NRR/AADB/BC NRR/SSI/BC NRR/EEEB/B C
OGC NRR/LPL2-1/BC NAME SLingam MOBrien RDennig MKotzalas MScott GWilson SHamrick EMarinos DATE 3/02/07 3/02/07 2/13/07*
12/18/06*
3/06/07 3/07/07 3/09/07 3/12/07 OFFICIAL RECORD COPY
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 250 Renewed License No. NPF-4 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al.,
(the licensee) dated October 3, 2006, as supplemented by letter dated January 24, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-4 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 250, are hereby incorporated in the renewed license. The
/licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-4 and the Technical Specifications Date of Issuance: March 13, 2007
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 230 Renewed License No. NPF-7 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company et al.,
(the licensee) dated October 3, 2006, as supplemented by letter dated January 24, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-7 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 230, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-7 and the Technical Specifications Date of Issuance: March 13, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 250 RENEWED FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 AND TO LICENSE AMENDMENT NO. 230 RENEWED FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Licenses and the Appendix "A" Technical Specifications (TSs) with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Pages Licenses Licenses License No. NPF-4, page 3 License No. NPF-4, page 3 License No. NPF-7, page 3 License No. NPF-7, page 3 TSs TSs 3.3.2-9 3.3.2-9 3.3.2-11 3.3.2-11 3.5.2-3 3.5.2-3 3.6.4-2 3.6.4-2 3.6.5-1 3.6.5-1 3.6.7-3 3.6.7-3 5.5-14 5.5-14
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 250 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 AND AMENDMENT NO. 230 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-338 AND 50-339
1.0 INTRODUCTION
By letter dated October 3, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML062850195), as supplemented by letter dated January 24, 2007 (ADAMS Accession No. ML070250058), Virginia Electric and Power Company (the licensee) submitted a request for changes to the North Anna Power Station, Unit Nos. 1 and 2 (North Anna 1 and 2), Technical Specifications (TSs). The licensee requested these TS changes as part of its resolution to the Nuclear Regulatory Commissions (NRCs) Generic Safety Issue (GSI) 191, Assessment of Debris Accumulation on PWR Sump Performance, and Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents at Pressurized-Water Reactor. Specifically, the licensee proposed to revise the method for starting the inside and outside recirculation spray (RS) pumps in response to a design-basis accident (DBA). Currently, the North Anna 1 and 2 RS pumps start by using delay timers that are initiated when the containment pressure reaches the containment depressurization actuation (CDA) High-High setpoint. The licensees proposed change would result in starting the RS pumps by a coincident CDA High-High pressure and refueling water storage tank (RWST) level Low. In addition, the proposed changes would replace the current LOCTIC containment methodology with the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) methodology described in Topical Report DOM-NAF-3, Revision 0, "GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment," (ADAMS Accession No. ML062420511) dated August 30, 2006.
The supplemental letter dated January 24, 2007, provided clarifying information that did not change the scope of the original application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination.
2.0 REGULATORY EVALUATION
Regulatory evaluation for the proposed TS changes consisted of the review of containment re-analysis using the methodology presented in the NRC staff-approved Topical Report DOM-NAF-3.
The NRC staff evaluated the radiological consequences of postulated DBAs after implementation of the proposed TS changes against the dose criteria specified in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.67.
Implementation of the alternative source term (AST) for analyzing DBAs was previously reviewed and approved for Renewed Facility Operating License Nos. NPF-4 and NPF-7 by License Amendment Nos. 240 and 221, dated June 15, 2005 (ADAMS Accession No. ML051590510), respectively. These amendments addressed the impact of the proposed TS changes on previously analyzed DBA radiological consequences and the acceptability of the revised analysis results. Because the loss-of-coolant accident (LOCA) analysis is the only DBA to take credit for the operation of the RS system for dose mitigation, it is also the only DBA analysis that is affected by the proposed TS changes.
The general design criteria (GDC) included in Appendix A to 10 CFR Part 50, did not become effective until May 21, 1971. The Construction Permits for North Anna 1 and 2 were issued prior to May 21, 1971; consequently, these units were not subject to GDC requirements (Ref.
SECY-92-223, dated September 18, 1992, ADAMS Accession No. ML003763736). The licensee stated that, however, the plant was designed to meet the intent of the draft GDC, including the following:
Criterion 38Containment heat removal which states that 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.
Criterion 50Containment design-basis which states that 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 loss-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 10 CFR 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.
The regulatory requirements upon which the NRC staff based its review, are Standard Review Plan (SRP) 15.0.1 and the accident dose criteria in 10 CFR 50.67, and the guidance in Regulatory Position 4.4 and Table 6 of Regulatory Guide (RG) 1.183. The licensee has not proposed any deviation or departure from the guidance provided in RG 1.183. The NRC staffs evaluation is based upon the following regulatory codes, guides, and standards:
10 CFR Part 50, Section 50.67, Accident source term.
10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants:
Criterion 19, Control room.
RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.
SRP Section 6.2.1, Containment Functional Design.
SRP Section 6.2.1.1.A, PWR Dry Containments, Including Subatmospheric Containments.
SRP Section 6.2.2, Containment Heat Removal Systems.
SRP Section 6.4, Control Room Habitability.
SRP Section 6.5.2, Containment Spray as a Fission Product Cleanup System.
SRP Section 15.0.1, Radiological Consequence Analysis Using Alternative Source Terms.
By a letter dated January 31, 2006, the licensee requested similar amendments for its Surry Power Station, Unit Nos. 1 and 2. On October 12, 2006, the NRC issued a safety evaluation (ADAMS Accession No. ML062920499) accepting the licensees request for Renewed Facility Operating License Nos. DPR-32 and DPR-37 by License Amendment Nos. 250 and 249, respectively.
3.0 TECHNICAL EVALUATION
North Anna 1 and 2 is a three-loop Westinghouse pressurized water reactor with a subatmospheric containment design. The engineered safeguards features that mitigate a LOCA or main steamline break accident (MSLB) event include the following (Chapters 5 and 6 of the North Anna 1 and 2 Updated Final Safety Analysis Report (UFSAR)):
A safety injection (SI) system that injects borated water into the cold legs of all three reactor coolant loops.
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 spray system, quench spray (QS) and RS, that operate together to reduce the containment temperature, return the containment pressure 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 CDA High-High containment pressure signal and draw suction from the 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. The RS pumps are started currently using delay timers that are initiated on the CDA High-High signal.
Each RS train has a recirculation spray heat exchanger that is cooled by SW (on the tube side) for long-term containment heat removal. The SI system consists of two LHSI and three high-head safety injection (HHSI) pumps that draw from the RWST and inject into the reactor coolant system 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 suction from the RWST to the discharge header of the LHSI pumps.
Because the RS and SI systems use the containment sump to show that design criteria are satisfied, the resolution of NRC GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (PWRs), affects the operation of IRS, ORS and LHSI pumps. Appendix A to RG 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Revision 3, dated November 2003, gives different criteria for showing 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 the net positive suction head (NPSH) margin. For a partially submerged strainer, the strainer debris head loss must be less than one-half the pool height (the Nuclear Energy Institute report, NEI-04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," (Ref. 1) also recommends the same criteria).
Currently, the North Anna 1 and 2 RS pumps start using delay timers that are initiated when the containment pressure reaches the CDA High-High 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 licensee proposed delaying the RS pump start until sufficient water level is available in the containment.
The licensee proposed to start the IRS and ORS pumps on 60 percent RWST wide range (WR) level coincident with a CDA High-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.
The current North Anna 1 and 2 licensing basis analysis methodology for LOCA containment response is the Stone & Webster LOCTIC computer code that is described in North Anna 1 and 2 UFSAR Chapters 5 and 6. The licensee proposed to replace the LOCTIC methodology with GOTHIC analytical methodology that is described in Topical Report DOM-NAF-3, GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment. In a letter dated November 1, 2005, the licensee submitted DOM-NAF-3 to the NRC. On August 30, 2006, the NRC issued a safety evaluation on this topical report accepting the licensees GOTHIC containment analysis methodology.
The GOTHIC analyses in the licensees October 3, 2006, submittal, are to replace the LOCTIC analyses in North Anna 1 and 2 UFSAR Chapters 5 and 6 for calculation of the following containment design requirements:
LOCA peak containment pressure and temperature, LOCA containment depressurization time, LOCA containment peak pressure following depressurization, NPSH available for the LHSI pumps, and NPSH available for the ORS and IRS pumps.
The licensee also used the minimum containment water level and maximum sump liquid temperatures from GOTHIC NPSH calculations to establish bounding inputs to the sump strainer design.
3.1 Containment Analysis In the letter dated October 3, 2006, the licensee proposed to revise its North Anna 1 and 2 containment analyses by converting from present Stone and Webster LOCTIC computer code to GOTHIC computer code. On August 30, 2006, the NRC issued a safety evaluation accepting Topical Report DOM-NAF-3, titled GOTHIC Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment, (ADAMS Accession No. ML062420511).
The licensee used the GOTHIC code to perform containment analyses for the current RS system configuration with delay timers (400 seconds for IRS pumps, 210 seconds for ORS pumps) and the current TS Figure 3.6.4-1, containment air partial pressure limits, and compared the results with LOCTIC analyses. The licensee concluded that while some design inputs have changed from the LOCTIC analyses, transient behavior was similar to the LOCTIC UFSAR analyses.
The licensee used the GOTHIC code to perform containment analyses for the proposed configuration by making two changes in the above analysis. One change was to assume that the RS pumps started at 60 percent RWST water level coincident with a CDA High-High containment pressure signal. The ORS pumps were assumed to start directly from the signal while IRS pumps were assumed to start 120 seconds after receiving the actuation signal. The licensee included instrument uncertainty for the level signal and the timer setpoint. The other change was to increase the TS containment air partial pressure used in the analysis to the values in the proposed TS Figure 3.6.4-1, as shown in Figure 1 at the end of this report. These analyses were to show that adequate margins to the containment acceptance criteria from the following North Anna 1 and 2 UFSAR existed:
LOCA and MSLB containment peak pressure < 45 psig LOCA containment pressure < 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 /> 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 EF LHSI Pump NPSH available > NPSH required ORS Pump NPSH available > NPSH required IRS Pump NPSH available > NPSH required.
3.1.1 Application of the GOTHIC Methodology Section 3.1 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the application of the GOTHIC methodology. The licensee stated that it used the containment response analysis methodology described in Topical Report DOM-NAF-3 without modification for the GOTHIC analyses.
The licensee described modeling of the geometry, engineered safeguards features, containment passive heat sinks, plant parameter design inputs, containment initial conditions and instrument uncertainty, and NPSH available and water holdup. Table 3.1-1 of to the licensees letter, dated October 3, 2006, provided key parameters used in the containment analysis. The licensee used conservative assumptions consistent with DOM-NAF-3. For example, the licensees assumptions in modeling NPSH available and water holdup for the LHSI, IRS and ORS pumps included the following:
Used a multiplier of 1.2 to the direct diffusion layer model heat transfer coefficients for passive heat sinks. This lowers the prediction for the containment pressure by transferring more energy from the containment atmosphere to the passive heat sinks, and thus, conservatively lowers the predicted NPSH available.
Assumed that all of the spray water is injected as droplets into the containment atmosphere (nozzle spray flow fraction of 1). This lowers the prediction for the containment pressure by transferring more energy from the containment atmosphere to the spray water, and thus, conservatively lowers the predicted NPSH available.
Used the upper limit on containment free volume. This lowers the prediction of the containment pressure, and thus, conservatively lowers the predicted NPSH available.
Used the minimum initial containment air pressure. This lowers the prediction of the containment pressure, and thus, conservatively lowers the predicted NPSH available.
Used a minimum sump pool surface area for the containment volume liquid/vapor interface area. This minimizes the heat transfer from the sump water to the containment atmosphere because during recirculation sump pool water is hotter than the containment atmosphere. Therefore, this assumption results in predicting a higher sump water temperature and a lower containment pressure, and thus, a conservatively lower NPSH available.
In Section 3.1.4 of Attachment 1 to its letter, dated October 3, 2006, the licensee stated that the surface area for metal heat sinks was revised based on a revised inventory that was documented in an internal calculation. The heat sink surface area used in the GOTHIC analysis was reduced by 5 percent from the sum of the calculated metal and concrete heat sink nominal surface areas. Noting that a minimum surface area would give conservative results for the containment peak pressure calculation, while the maximum surface area would give conservative results for the available NPSH calculation, the NRC staff requested the licensee to explain and justify what area was used for the calculations. In response, in a letter dated January 24, 2007, the licensee stated that minimum heat sink surface area is conservative and was used for both calculations for the following reasons. Sensitivity studies performed by the licensee showed that the minimum available NPSH for these pumps was mostly insensitive to the heat sink surface area. The licensee explained that passive heat sinks absorb energy early in the containment heatup but return energy to the containment atmosphere as the spray systems depressurize and cool the containment. When the minimum available NPSH occurs, the containment is depressurized below atmospheric pressure, the total energy transferred to the heat sinks is the same whether the lower or higher value of the heat sink surface area is used. In the GOTHIC analysis, the 5 percent less than nominal value of the heat sink surface area produced a slightly lower minimum available NPSH. The NRC staff considers the licensees justification acceptable.
The NRC staff determined that the licensees GOTHIC analyses are consistent with the NRC staff-accepted Topical Report DOM-NAF-3 and is appropriate for analyzing the proposed changes.
3.1.2 Break Mass and Energy Releases Section 3.2 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the modeling of the break mass and energy releases following a LOCA or MSLB. For the LOCA, the licensee applied the break release methodology in Section 3 of DOM-NAF-3. For the MSLB, the licensee conservatively used North Anna 1 and 2 mass and energy release data from WCAP-11431, which were generated using the NRC-approved methodology from WCAP-8822 (Refs. 2 and 3, respectively).
The NRC staff determined that the modeling of the break mass and energy releases were appropriate for GOTHIC containment analyses because the licensee used NRC-accepted methodology.
3.1.3 LOCA Peak Pressure and Temperature Section 3.3 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the calculation of the LOCA peak pressure and temperature. The LOCA peak containment temperature is related to the peak pressure because the containment atmosphere is saturated when the peak pressure occurs.
Section 2.4 of Attachment 1 to the licensees letter, dated October 3, 2006, states the current configuration containment TS maximum air temperature of 120 EF as the limiting initial temperature for LOCA analyses. For the new configuration, the licensee has proposed to reduce this limit to 115 EF in order to increase pump NPSH margin. The NRC staff requested the licensee to confirm that the existing containment ventilation system has sufficient capacity to maintain the containment air temperature less than or equal to the proposed TS maximum limit of 115 EF. In response, in the letter dated January 24, 2007, the licensee confirmed that the ventilation system was designed to maintain the containment air temperature less than 105 EF which is conservative. The NRC staff considers licensees response acceptable.
The licensees GOTHIC analysis results show that a double-ended hot-leg guillotine (DEHLG) break causes a more limiting blowdown peak pressure than the double-ended pump suction guillotine (DEPSG) break. GOTHIC code predicted less peak temperature (269.3 EF) for DEHLG break for proposed configuration as compared to current configuration (269.8 EF). The GOTHIC code prediction for the peak pressure for the proposed configuration was higher than that for the current configuration (57.4 psia versus 56.8 psia for DEHLG break). This resulted from a higher TS containment air partial pressure assumed for the initial condition for the proposed configuration than the current configuration (12.3 versus 11.7 psia). The delay proposed for the starting of RS pumps would not affect the actual LOCA peak pressure and temperature which occur much earlier than the start of RS pumps.
For both current and proposed configurations, the containment peak pressure predicted is less than the containment design limit of 59.7 psia. The containment vapor temperature and liner temperature predicted were below the design limit of 280 EF.
Using acceptable conservative calculations, the licensee determined that the proposed maximum operating containment air partial pressure of 12.3 psia gives the LOCA peak pressure and temperature below the design limits. The NRC staff finds the licensees LOCA peak pressure and temperature acceptable.
3.1.4 LOCA Containment Depressurization Section 3.4 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the calculation of the 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 the LOCTIC code show that the containment is subatomospheric within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and remains subatmospheric thereafter.
The DEPSG break is limiting because it has the largest energy release to the containment due to the available energy removal from the steam generator secondary side. The licensee performed GOTHIC analyses to determine containment depressurization time (CDT) and the depressurization peak pressure (DPP) for a DEPSG break for both current and proposed configurations. The CDT represents the time it takes for the containment pressure to first drop below atmospheric pressure after the pipe break. For the current configuration, the containment pressure reached subatomospheric conditions at 2,604 seconds (i.e., CDT) and the DPP was -1.06 psig at 6,077 seconds. Therefore, the current configuration maintains a subatmospheric containment after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. For the proposed configuration, the licensee performed several sensitivity analyses to identify the limiting CDT and DPP results. Of these the limiting case involved the proposed TS maximum air partial pressure of 12.3 psia and SW temperature of 55 EF which gave a limiting CDT of 3,205 seconds (time for containment pressure to reach less than 2 psig) and limiting DPP of 0.78 psig at 6,680 seconds. The case which gave the limiting time of 14,530 seconds for the containment pressure to reach subatmospheric was with TS maximum air partial pressure of 10.4 psia and SW temperature of 95 EF. These values are below the containment leak rate assumption values used for the dose analysis (Table 4.1-1 of Attachment 1 to the licensees letter, dated October 3, 2006): pressure decreasing to 2.0 psig from 0 to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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 />, and subatmospheric after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Using acceptable conservative calculations, the licensee determined LOCA containment depressurization parameters, CDT and DPP. Based on its review, the NRC staff finds the licensees calculation of CDT and DPP acceptable.
3.1.5 LHSI Pump NPSH Analysis Section 3.5 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the LHSI pump NPSH analysis. The GOTHIC analysis predicted the LHSI Pump NPSH margin for the current and proposed configuration limiting cases. The current configuration limiting case involved TS initial containment air partial pressure of 9.0 psia and SW temperature of 95 EF.
The proposed configuration limiting case involved TS initial containment air partial pressure of 10.3 psia and SW temperature of 75 EF. The LHSI pump NPSH margin predicted for the current and proposed configuration limiting cases were 1.09 ft and 1.57 ft of water respectively, for a required NPSH of 13.4 ft of water.
Using acceptable conservative GOTHIC calculations, the licensee confirmed that the LHSI pump will perform its intended function. Based on its review, the NRC staff finds the licensees analysis acceptable.
3.1.6 RS Pump NPSH Analysis Section 3.6 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the RS pump NPSH analysis which was performed using the GOTHIC code.
The licensee analyzed the current and proposed configurations for the RS pump NPSH using GOTHIC methodology. For the current configuration, the licensee analyzed the DEHLG break, which is limiting because the blowdown energy data maximize the energy in the containment sump early in the accident which raises the sump temperature. For the proposed configuration with delayed RS pump start, the licensee analyzed both DEHLG and DEPSG breaks for a range of single failures cases: (a) emergency bus, (b) QS pump, © casing cooling pump, (d) LHSI pump, (e) ORS pump, and (f) IRS pump.
For the current configuration, the licensee calculated IRS and ORS pump NPSH available of 12.17 ft and 15.3 ft of water respectively, giving corresponding NPSH margins of 2.57 ft and 4.0 ft of water. For the proposed configuration, for IRS pump, the licensee found that the limiting case was a DEPSG break in conjunction with a single failure of an emergency bus; the calculated minimum available NPSH was 15.12 ft giving a margin of 5.52 ft of water. For the ORS pump, the licensee found that the limiting case was a DEPSG break in conjunction with single failure of a casing cooling pump; the calculated minimum available NPSH was 18.73 ft giving a margin of 7.43 ft of water.
Insert No. 9 of Attachment 4 to the licensees letter, dated October 3, 2006, states that when the coincident logic of RWST level-low with containment pressure high-high is satisfied, an automatic signal starts the ORS pump immediately, and the IRS pump after 120-second time delay. The NRC staff requested the licensee to explain how would the logic verify the availability of sufficient water level in the sump before starting the IRS pump. In response, in letter dated January 24, 2007, the licensee stated that IRS pump automatic start logic does not use containment sump level as an input. For all design-basis LOCAs, the GOTHIC analyses confirm that sufficient water volume will reach the containment sump to meet the RS strainer submergence and NPSH requirements before the IRS pump starts. The GOTHIC analyses include conservatisms for the parameters that affect the containment sump water level, such as water held inside the containment that would not reach the containment sump, initial RWST water level, and uncertainty in the RWST water level measurement. Considering all design-basis LOCAs, the licensee has calculated the minimum containment sump water level of 1.86 ft assuming a conservative holdup volume in containment of about 42,400 gallons and a 2.5 percent uncertainty in the RWST water level measurement. The NRC staff considers the licensees response acceptable.
Using acceptable conservative calculations, the licensee confirmed that the RS pump will perform its intended function. Based on its review, the NRC staff finds the licensees analysis acceptable.
3.1.7 MSLB Peak Pressure and Temperature Section 3.7 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the MSLB peak pressure and temperature analyses.
The GOTHIC analyses for the MSLB do not credit the recirculation spray system. This precludes the need to show adequate RS pump performance during an 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.
For the current and proposed configurations, the GOTHIC MSLB analyses predicted containment peak pressure of 57.84 psia and 57.65 psia. These values are more limiting than those predicted in the LOCA analyses described in Section 3.1.3 of this report.
For current and proposed configurations, the licensee calculated the MSLB peak containment air temperature of 318.4 EF and 308.4 EF respectively, which are higher than the design limit of 280 EF. The licensees analyses included an additional 1 ft2 thermal conductor to determine 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 direct diffusion layer model heat transfer coefficient. The peak liner temperature for the proposed configuration was 258 EF, and therefore, the MSLB peak containment air temperature exceeding the containment design limit did not adversely affect the containment liner. Section 3.1.10 of this report describes the effect and concludes that the MSLB peak containment air temperature exceeding the design limit will not adversely affect the safety-related equipment inside the containment.
Using acceptable conservative calculations, the licensee calculated (1) MSLB peak containment pressure, which is bounded by the containment design peak pressure and (2) MSLB peak containment air temperature, which exceeded the containment design temperature. The licensee showed that the MSLB peak containment air temperature exceeding design limit did not adversely affect the containment liner or the safety-related equipment inside the containment. Based on its review, the NRC staff finds the licensees analysis acceptable.
3.1.8 Inadvertent QS Actuation Event Section 3.8 of Attachment 1 to the licensees letter, dated October 3, 2006, describes the inadvertent QS actuation analysis. North Anna 1 and 2 UFSAR Section 6.2.6.3 describes the containment response for an inadvertent QS actuation analysis. The licensee proposed to increase the TS containment air partial pressure minimum limit to 10.3 psia, as described in Section 3.8 of Attachment 1 to the licensees letter, dated October 3, 2006. The internal pressure capability limit that the containment liner can withstand is 5.5 psia minimum as per North Anna 1 and 2 UFSAR Section 6.3.6.3. The licensees analysis shows that for an inadvertent QS actuation at the TS minimum containment air partial pressure (10.3 psia) and without crediting for an operator action to terminate the spray, the containment internal pressure is 8.62 psia. This is greater than the 5.5 psia minimum containment liner limit, and, therefore, meets the liner design criteria. The NRC staff finds the licensees analysis acceptable.
3.1.9 Impact on Emergency Diesel Generators The licensee was requested to review the revised loading sequence of the emergency diesel generators (EDG) with the ORS and IRS pumps along with other engineered safety featured loads and verify that it meets the position 4 of Safety Guide 9 which states: At no time during the loading sequence should the frequency and voltage decrease to less than 95 percent of nominal and 75 percent of nominal, respectively. During recovery from transients caused by step load increases or resulting from disconnection of the largest single load, the speed of the diesel generator set should not exceed 75 percent of the difference between nominal speed and the overspeed trip setpoint or 115 percent of nominal, whichever is lower. Voltage should be restored to within 10 percent of nominal and frequency should be restored to within a 2 percent of nominal in less than 40 percent of each load sequence time interval. In letter dated January 24, 2007, the licensee confirmed that the existing dynamic load analysis for the EDG remains bounding for the proposed change based on analysis of the predicted system response during a LOCA with a loss of offsite power.
3.1.10 Equipment Qualification (EQ)/EQ Envelope Verification In its submittal dated October 3, 2006, the licensee stated that it had developed new pressure and temperature EQ envelopes that sufficiently bounded the GOTHIC LOCA and the MSLB pressure and temperature profiles. By letter dated January 24, 2007, the licensee indicated that the composite pressure and temperature profiles were developed from the LOCA and MSLB pressure and temperature profiles from the NRC staff-approved GOTHIC methodology.
These composite profiles were then compared to the EQ test reports for all environmentally qualified equipment inside the containment. As a result of this comparison, the licensee concluded that the environmentally qualified equipment inside the containment was qualified for the GOTHIC accident analysis profiles for pressure and temperature.
Therefore, based on the licensees assessment documented in the submittal on October 3, 2006, the NRC staff concluded that the GOTHIC accident profile is enveloped by the EQ test report. The EQ status of equipment inside containment is not affected by the revised containment temperature and pressure profiles resulting from delaying RS pump start and increasing the containment air partial pressure limits. The NRC staff finds that the licensees EQ envelope verification is acceptable and complies with 10 CFR 50.49.
3.1.11 Proposed TS Limits for Containment Air Partial Pressure versus SW Temperature In Section 3.10 of Attachment 1 to its letter, dated October 3, 2006, the licensee describes the proposed TS limits for containment air partial pressure versus SW temperature as given in TS Figure 3.6.4-1, which is shown in Figure 1 of this report.
This operating domain maintains the current limits of 35 EF to 95 EF for SW temperature but reduces the maximum containment air temperature from 120 EF to 115 EF, and accounts for 0.30 psi instrument uncertainty for the containment air partial pressure. The following defined the allowable containment air partial pressure limits as shown in Figure 1 of this report:
The containment depressurization analyses (Section 3.1.4) limit the maximum operating air partial pressure to 12.3 psia at 55 EF SW temperature. The LOCA peak pressure analysis (Section 3.1.3) and the MSLB peak pressure analysis (Section 3.1.7) show a margin to the containment design limit of 45 psig. Therefore the TS limit is maintained constant at 12.3 psia from 35 EF to 55 EF SW temperature.
The containment depressurization analyses (Section 3.1.4) set the TS upper limit from 12.3 psia at 55 EF SW to 10.4 psia at 95 EF SW. The allowable air partial pressure decreases with increasing service water temperature because it is more difficult to depressurize the containment at higher service water temperature. To meet subatmospheric requirements, the initial air partial pressure is limited to 10.4 psia at 95 EF service water.
The LHSI pump NPSH analyses (Section 3.1.5) set the lower limit on air partial pressure (the RS pumps use the same assumptions but have more NPSH margin (Section 3.1.6)). The proposed lower limit in Figure 1 ensures at least 1.5 ft of NPSH margin across the entire SW temperature range.
The NRC staff determined that the licensees analyses described in Sections 3.1.3 through 3.1.7 of this report provide bases for the proposed changes to TS Figure 3.6.4-1.
The licensee included in its application the revised TS Bases to be implemented with the TS changes. The NRC staff finds that the TS Bases Control Program is the appropriate process for updating the affected TS Bases pages and has, therefore, not included the affected Bases pages with this amendment.
3.1.12 Summary of Containment Analysis Results In Table 3.11-1 of Attachment 1 to its letter, dated October 3, 2006, which is given as Table 1 in this report, the licensee summarized the containment analysis results and compared them to the design limits.
The results of the GOTHIC analyses for the proposed configuration show that all containment analysis acceptance criteria, except for the MSLB peak temperature are met for operation in the allowable region of Figure 1 and starting the RS pumps on 60 percent RWST WR level coincident with CDA High-High containment pressure. See Sections 3.1.7 and 3.1.10 of this report for the disposition of MSLB peak temperature exceeding the design limit. GOTHIC calculated MSLB temperatures, which are greater than 280 EF, do not adversely affect the integrity of the containment liner.
3.2 Radiological consequences of the design basis LOCA The NRC staff reviewed the regulatory and technical analyses performed by Dominion in support of its proposed license amendment, as they relate to the radiological consequences of the design basis LOCA. Information regarding these analyses was provided in Attachment 1 of the October 3, 2006, submittal. Additionally, the containment re-analysis presented in Topical Report DOMNAF-3, Revision 0, Gothic Methodology for Analyzing the Response to Postulated Pipe Ruptures Inside Containment, dated October 2005, and referenced in the licensees analysis, was previously reviewed and approved by the NRC staff in a separate safety evaluation (ADAMS Accession No. ML062090464).
The NRC staff also reviewed all other assumptions, inputs, and methods used by Dominion to assess the impacts of the proposed license amendment. The NRC staff performed independent calculations to confirm the conservatism of the licensees analyses.
To perform independent confirmatory dose calculations for the LOCA, the NRC staff used the NRC-sponsored radiological consequence computer code, RADTRAD: Simplified Model for RADionuclide Transport and Removal And Dose Estimation, Version 3.03, as described in NUREG/CR-6604, A Simplified Model of Aerosol Removal by Containment Sprays, dated June 1993. The RADTRAD code, developed by the Sandia National Laboratories for the NRC, estimates transport and removal of radionuclides and the resulting radiological consequences at selected receptors.
3.2.1 Loss-of-Coolant Accident (LOCA)
With respect to evaluating the LOCA dose, the proposed TS changes will result in the post-LOCA start of the RS pumps, which mitigate activity release, with a coincident containment depressurization actuation (CDA) High-High pressure and RWST Level Low. Currently, RS pumps are started using delay timers that are initiated after containment pressure has reached the CDA High-High setpoint. This proposed method for starting RS pumps will delay activation of the RS system, which will result in a short-term increase in air leakage from the containment and a short-term reduction in spray removal of airborne radioactivity from the containment atmosphere.
To reflect the implementation of the new RS pump start methodology, and to offset the potential increase in calculated radiological consequences, the licensee has proposed changes to the current design basis LOCA analysis, which are detailed in the following subsections.
3.2.1.1 Revised Assumptions - RS Pump Restart The following two changes to the LOCA analysis assumptions result from the implementation of the new recirculation spray start methodology discussed in Topical Report DOM-NAF-3, and approved by the NRC staff in a safety evaluation dated August 1, 2006 (ADAMS Accession No. ML062090464):
The RS system operation for spray removal is delayed from 288.5 seconds to 40 minutes.
The early Outside RS (ORS) pump start occurs at 14 minutes for emergency core cooling system (ECCS) leakage versus 288.5 seconds in the current basis.
Because the RS pump start methodology of DOM-NAF-3 was found acceptable by the NRC staff in the above-mentioned safety evaluation, these changes to the assumptions of the LOCA analysis are justified and also found by the NRC staff to be acceptable.
3.2.1.2 Revised Assumptions - Containment Analysis In addition to the proposed implementation of the DOM-NAF-3 RS pump start methodology, the licensee proposes to revise the post-LOCA containment analysis of temperature and pressure response based on the GOTHIC computer code modeling. The following LOCA analysis assumptions are proposed based on the implementation of this new containment analysis:
Containment leakage after the first hour of a LOCA is increased to 0.04 percent-volume-per day for the time period of 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> versus 0.021 percent-volume-per-day for the time period of 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. The newly calculated containment depressurization profile, decreases to 2.0 psig, after first hour.
The containment sump volume varies as a function of time and is increased by 10 percent for added conservatism.
RWST backleakage is assumed to start at 31.8 minutes vs. 30 minutes in the current licensing basis. This is also consistent with the start of Recirculation Mode Transfer (RMT).
The NRC staff finds these assumptions, which are based on the methodology detailed in Topical Report DOM-NAF-3, to also be acceptable for the LOCA analysis.
3.2.1.3 Revised Assumptions - Spray Removal Modeling Because of the proposed changes to the RS pump start methodology, the licensee also affected the timing and activation characteristics of the containment spray systems. Therefore, assumptions relating to the calculation of aerosol activity removal in containment were modified to reflect these changes. The following revised assumptions are proposed:
Use specific spray volumes for each of the Quench Spray (QS) only operation, the combined QS/RS operation, and the RS only operation, versus using 1 sprayed volume for the entire period of spray operation.
Newly calculated changes in aerosol removal coefficients due to the delay in RS operation and conservative QS flow rate assumptions.
These two assumptions are revised based on methodology described in NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays. This spray removal modeling described in this NUREG is an acceptable methodology for calculating aerosol removal by sprays, as discussed in RG 1.183. So, these assumptions, based on approved methodology, are acceptable.
3.2.1.4 Revised Assumptions - Other Assumptions The following table lists additional revised assumptions, and the justification for their acceptability, the licensee proposes to make to the LOCA analysis:
Proposed Revised LOCA Assumptions Justification for Acceptability 1
Credit for 96 and 720-hour CR Occupancy Factor reduction.
Consistent with RG 1.183 guidance.
2 Credit for timed release of radionuclides into containment sump.
Consistent with RG 1.183 guidance.
3 Increase of Decontamination Factor (DF) for RWST releases from 10 to 40.
Calculation of DF is consistent with conservatively limited SRP 6.5.2 equation.
4 Increase of assumed containment volume to 1.916E+06 ft3.
Conservative adjustment to the value used for the current DBA LOCA analysis (ML032670821).
5 Increase in assumed auxiliary building filter efficiency for organic iodines from 70% to 90%.
Consistent with current TS filter surveillance requirements.
6 Increase in assumed control room filter efficiency for organic iodines from 70% to 95%.
Consistent with current TS filter surveillance requirements.
In addition to the revised assumptions listed in the table above, the licensee also recalculated the North Anna 1 and 2 control room volume and reduced their RWST breathing rate, i.e.,
release rate, from a rounded value of 4 cfm to the actual value of 3.7 cfm.
3.2.2 Airborne Radioactivity Removal by Containment Sprays The containment depressurization spray system at North Anna 1 and 2 consists of two separate spray subsystems, the QS and RS. These two subsystems operate in parallel to reduce the containment temperature, return the containment pressure to subatmospheric, and remove heat from containment. At an established time during the accident, the RS operates alone to maintain the containment subatmospheric and transfer heat from the containment to the service water (SW) system.
The QS system consists of two pumps that start on a CDA High-High containment pressure signal; this will remain unaffected by the proposed changes. 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. The LOCA analysis assumes a failure of one train of engineered safeguards equipment; therefore, only one QS train and one RS train, for each of its IRS and ORS subsystems, is credited. This is discussed in the current LOCA analysis, and found to be acceptable in the NRC staffs safety evaluation dated June 15, 2005 (ADAMS Accession No. ML051590510).
The proposed changes will start the IRS and ORS pumps on a 60-percent RWST wide range level, coincident with a CDA High-High pressure signal. The ORS will start immediately once this logic is satisfied, while the IRS will start with a 120-second delay timer upon this logic being satisfied. Equipment associated with accomplishing this spray actuation has been classified by Dominion as part of the engineered safety features actuation system (ESFAS), and identified as safety grade where applicable.
The revised containment analysis, approved by the NRC staff in a separate safety evaluation (ADAMS Accession No. ML062090464), provided the timing of QS and RS operation, as well as the spray flow rates corresponding to the calculated rate of containment depressurization.
The licensee credits airborne radioactivity removal by the QS only until its termination at 1.50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. After that, the airborne radioactivity removal is credited by the RS only, until its cessation at 7.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
The licensee calculated the aerosol spray removal coefficients using a model described in NUREG/CR-5966. This model provides a calculational method for determining aerosol (particulate) activity removal that is acceptable to the NRC staff, as noted in RG 1.183. The removal rate was calculated based on the 10th percentile level equation, which the NRC staff considers to be conservative, and, therefore, acceptable.
During a certain post-accident time period, both the QS and RS systems are postulated to be in operation. At North Anna 1 and 2, this occurs at 0.667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br /> following accident initiation and continues to 1.50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Analytically, this condition will result in the combination of radioactivity removal coefficients associated with both systems. The methodology used by the licensee to combine these removal coefficients employs a weighted averaging of system characteristics; i.e., the spray heights that individually characterize the QS and RS systems are weighted by the respective system spray flow, over the combined total spray flow. Because it reflects a conservative implementation of the available guidance, and avoids a simple addition of calculated removal coefficients, the NRC staff finds this methodology to be acceptable for combining the removal associated with the two containment spray systems. The combined QS and RS removal coefficients, as submitted in the amendment request, are given in Table 2.
For elemental iodine removal by sprays, a removal coefficient of 10 hr-1 was used, as is consistent with approved methodology shown in the North Anna 1 and 2 UFSAR. This value is conservative and found acceptable for this re-analysis.
3.2.3 Atmospheric Dispersion The licensee used atmospheric dispersion factors (/Q values) from their current licensing basis.
For the control room (CR), exclusion area boundary (EAB), and low population zone (LPZ) dose estimates, the /Q values implemented in the amendment request were previously reviewed and approved by the NRC staff as part of the licensees request to implement the AST methodology (ADAMS Accession No. ML051590510).
3.2.4 Summary of Revised LOCA Analysis The licensee evaluated the radiological consequences resulting from the postulated LOCA and concluded that the CR, EAB, and LPZ radiological consequences are within the dose acceptance criteria provided in 10 CFR 50.67 and the accident dose guidelines specified in SRP 15.0.1. The NRC staffs review has found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this safety evaluation. The NRC staff performed independent analyses of radiological dose consequences using the licensees assumptions and confirmed the licensees results. The NRC staff finds that the CR, EAB, and LPZ dose consequences, estimated by the licensee for the LOCA, as presented in Table 3, meet the dose acceptance criteria in 10 CFR 50.67 of 25 rem Total Effective Dose Equivalent (TEDE) at the EAB and LPZ, and 5 rem TEDE in the CR, and are, therefore, acceptable.
4.0
SUMMARY
The licensee proposed operation in the allowable region of Figure 1 and starting the RS pumps on 60 percent RWST WR level coincident with CDA High-High containment pressure. The licensee showed using the NRC-approved GOTHIC Methodology for Analyzing the Response to Postulated Ruptures Inside Containment, that the proposed changes would not affect the containment analyses acceptance criteria (NRC staff safety evaluation, dated August 30, 2006).
The NRC staff determined that the proposed changes meet the intent of (1) GDC 38 because the licensee showed that the containment sprays would remove containment heat to reduce containment pressure and temperature rapidly, consistent with the functioning of other associated systems, following design-basis LOCA and would maintain them at acceptably low levels and (2) GDC 50 because the licensee showed that the reactor containment structure, including access openings, penetrations, and the containment heat removal system can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from design-basis LOCA.
As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of a postulated LOCA with the proposed TS changes. The NRC staff finds that the licensee used analysis methods and assumptions consistent with the regulatory requirements and guidance identified in Section 2.0. The NRC staff compared the doses estimated by the licensee to the applicable criteria, and finds, with reasonable assurance, that the licensees estimates of the CR, EAB, and LPZ doses will comply with these criteria. Therefore, the NRC staff finds that the proposed license amendment is acceptable with respect to the radiological consequences of DBAs.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendments. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (71 FR 70563). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (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) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Table 1 GOTHIC Containment Analysis Results Acceptance Criterion Design Limit Current Configuration Proposed Configuration LOCA Peak Pressure 59.7 psia 56.8 psia 57.4 psia LOCA Peak Temperature 280 EF 269.8 EF 269.3 EF MSLB Peak Pressure 59.7 psia 57.54 psia 57.65 psia MSLB Peak Temperature 280 EF 318.4 EF(*)
308.4 EF(*)
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 /> or less
< 2.0 psig at 2081 sec
< 2.0 psig at 3205 sec Depressurization Peak Pressure
< 2.0 psig at 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
-1.06 psig at 6077 sec
+0.78 psig at 6680 sec LHSI Pump NPSH Available 13.4 ft at 4050 gpm 14.49 ft at 4050 gpm 14.97 ft at 4050 gpm IRS Pump NPSH Available 9.6 ft at 3400 gpm 12.17 ft at 3400 gpm 15.12 ft at 3400 gpm ORS Pump NPSH Available 11.3 ft at 3750 gpm 15.30 ft at 3750 gpm 18.73 ft at 3750 gpm
- See Sections 3.1.7 and 3.1.10 of this report for the disposition of MSLB peak temperature exceeding the design limit.
Table 2 Recalculated Aerosol Removal Coefficients versus Time Spray System in Operation Time Steps Removal
- Constant, From (hr)
To (hr)
(hr-1)
QS Only 2.03E-02 5.56E-01 5.832E+00 QS Only 5.56E-01 6.67E-01 6.167E+00 QS & RS 6.67E-01 8.33E-01 1.256E+01 QS & RS 8.33E-01 1.11E+00 1.267E+01 QS & RS 1.11E+00 1.39E+00 1.267E+01 QS & RS 1.39E+00 1.50E+00 1.256E+01 RS Only 1.50E+00 1.80E+00 1.214E+01 RS Only 1.80E+00 1.88E+00 7.739E+00 RS Only 1.88E+00 1.97E+00 5.406E+00 RS Only 1.97E+00 2.35E+00 2.885E+00 RS Only 2.35E+00 3.82E+00 1.569E+00 RS Only 3.82E+00 5.46E+00 1.402E+00 RS Only 5.46E+00 7.13E+00 1.380E+00 Table 3 Licensee Calculated DBA LOCA Radiological Dose Consequences Control Room EAB LPZ (rem TEDE)
Current 5.0**
1.85 0.12 Recalculated 4.1 2.1 0.2 10 CFR 50.67 Limits 5
25 25
Figure 1. Containment air partial pressure versus service water temperature (proposed TS Figure 3.6.4-1)
8.0 REFERENCES
1.
Technical Report NEI-04-07, Revision 0, Pressurized Water Reactor Sump Performance Evaluation Methodology, Volumes 1 and 2 (Safety Evaluation Report),
December 2004.
2.
WCAP-11431, Revision 0, Mass and Energy Releases Following a Steam Line Rupture for North Anna Units 1 and 2, February 1987.
3.
WCAP-8822-P, Mass and Energy Releases Following a Steam Line Rupture, September 1976, with Supplements 1 (WCAP-8822-S1-P-A) and 2 (WCAP-8822-S2-P-A) both dated September 1986 (WCAP-8860 is the Non-Proprietary version).
Principal Contributors: A. Sallman H. Wagage A. Boatright Date: March 13, 2007
North Anna Power Station, Units 1 & 2 cc:
Mr. David A. Christian Senior Vice President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrooks Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 Mr. C. Lee Lintecum County Administrator Louisa County Post Office Box 160 Louisa, Virginia 23093 Ms. Lillian M. Cuoco, Esq.
Senior Counsel Dominion Resources Services, Inc.
Building 475, 5 th floor Rope Ferry Road Waterford, Connecticut 06385 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23218 Old Dominion Electric Cooperative 4201 Dominion Blvd.
Glen Allen, Virginia 23060 Mr. Chris L. Funderburk, Director Nuclear Licensing & Operations Support Dominion Resources Services, Inc.
Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060-6711 Office of the Attorney General Commonwealth of Virginia 900 East Main Street Richmond, Virginia 23219 Senior Resident Inspector North Anna Power Station U. S. Nuclear Regulatory Commission P. O. Box 490 Mineral, Virginia 23117 Mr. Daniel G. Stoddard Site Vice President North Anna Power Station Virginia Electric and Power Company Post Office Box 402 Mineral, Virginia 23117-0402 Dr. Robert B. Stroube, MD, MPH State Health Commissioner Office of the Commissioner Virginia Department of Health Post Office Box 2448 Richmond, Virginia 23218