ML042440647
| ML042440647 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/18/2004 |
| From: | Grecheck E Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 04-494 | |
| Download: ML042440647 (9) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICIfINIoNI, VIRGINIA 23261 August 18, 2004 U.S. Nuclear Regulatory Commission Serial No.04-494 Attention: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (Dominion)
NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGES IMPLEMENTATION OF ALTERNATE SOURCE TERM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION In a letter dated September 12, 2003 (Serial No.03-464) Dominion requested amendments in the form of changes to the Technical Specifications to Facility Operating Licenses Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes were requested based on the radiological dose analysis margins obtained by using an alternate source term consistent with 10 CFR 50.67.
In an August 3, 2004 telephone conference call, the NRC Staff requested additional information regarding the dose analysis methods/assumptions, charcoal filter testing efficiencies, and the proposed Technical Specifications changes. Information regarding the dose analysis, including some corrected analysis input data (i.e., pages 50 and 68 of the original dose assessment) is included in the attachment to this letter.
During the phone call the NRC provided their position on the assumptions to be used for evaluating the radiological consequences of a fuel handling accident consistent with RG 1.183.
This position specifically addresses the use of effective decontamination factors (DFs) for fuel handling accidents. Based on the NRC position on the use of effective DFs instead of the DFs for elemental and organic species, additional dose assessment will be required to support the requested Technical Specification changes.
Therefore, the remainder of the requested information will be provided in a subsequent letter.
Due to the number of program and procedure changes necessary to implement these changes, we continue to request ninety days from the issuance date of the amendments to implement the Technical Specifications changes. If you have any further questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.
Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services A00 1
Letter No.04-494 Docket Nos. 50-338/339 Attachment 1.
Request for Additional Information and Revised Analysis Input Data Commitments made in this letter: None cc:
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060 Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, VA 23218 Mr. M. T. Widmann NRC Senior Resident Inspector North Anna Power Station Mr. S. R. Monarque NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8-H12 Rockville, MD 20852 2
SN: 04-494 Docket Nos.: 50-338/339
Subject:
AST RAI - Dose Analysis COMMONWEALTH OF VIRGINIA
)
)
COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck who is Vice President -
Nuclear Support Services 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 her knowledge and belief.
Acknowledged before me this.1 day of
, 2004.
My Commission Expires:
313)1 b
/ \\otary Public
- a-\\ (SOL)
Letter No.04-494 Docket Nos. 50-338/339 Attachment Attachment Proposed Technical Specification Changes Implementation of Alternate Source Term Response to Request for Additional Information and Revised Analysis Input Data pages 58 and 68 From September 12, 2003 letter (Serial No.03-464)
Virginia Electric and Power Company (Dominion)
North Anna Power Station Units 1 and 2
Letter No.04-494 Docket Nos. 50-338/339 Attachment NRC Question:
Provide the basis for reducing the ground level X/Q value calculated with ARCON96 by a factor of 5 at the end of the release phase for the steam generator tube rupture.
Dominion Response:
The North Anna safety analysis of a steam generator tube rupture (SGTR) accident models large quantities of radioactive steam flowing out of the steam generator power operated relief valves (PORVs). The radiological impact on the control room of this large steam release is determined with atmospheric dispersion factors.
From Regulatory Guide 1.194 Section 6, "Plume Rise", the excerpted paragraph below discusses the reduction of atmospheric dispersion factors (X/Q's) for releases from steam relief valves and atmospheric dump valves. The North Anna steam generator PORVs are examples of atmospheric dump valves.
"In lieu of mechanistically addressing the amount of buoyant plume rise associated with energetic releases from steam relief valves or atmospheric dump valves, the ground level X/Q value calculated with ARCON96 (on the basis of the physical height of the release point) may be reduced by a factor of 5. This reduction may be taken only if 1) the release point is uncapped and vertically oriented and 2) the time-dependent vertical velocity exceeds the 95th-percentile wind speed (at the release point height) by a factor of 5."
As was discussed in our letter dated November 20, 2003 (Serial No. 03-464A), the North Anna steam generator PORVs are uncapped and vertically oriented. The 95t-percentile wind speed at the PORV release height is 5.72 meters per second. Five times this value for wind speed is 28.6 meters per second. Therefore, if the vertical velocity of the steam exiting the steam generator PORVs after a SGTR accident is greater than 28.6 meters per second, the atmospheric dispersion factors used to model the PORVs can be reduced by a factor of five. In the letter dated November 20, 2003 (Serial No. 03-464A) Dominion provided the vertical steam velocities for the steam generator PORVs. After isolation of the affected steam generator, the vertical steam velocity of the unaffected steam generator PORVs was determined based on a conservatively low average steam release velocity instead of a minimum velocity.
However, to apply this factor of five reduction for the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release duration, the vertical velocity of the exiting steam must be greater than 28.6 meters per second at all times, even at the end of the accident when steam velocities are at their lowest.
During the first 30 minutes of a SGTR accident, initial plant cooldown, reactor coolant system (RCS) depressurization and secondary side isolation are expected to occur and are modeled in the safety analysis. Subsequent to the first 30 minutes of the SGTR accident, the emergency procedures direct the operators to proceed with plant
Letter No.04-494 Docket Nos. 50-338/339 Attachment cooldown.
The cooldown instructions limit the cooldown rate to <100 OF/hour.
Therefore, residual heat removal (RHR) initiation conditions (350 OF) would be expected to be reached in 2-3 hours.
To conservatively model both integrated steam releases and low steam PORV exit velocity for X/Q considerations, a slow cooldown is modeled with 350 OF being reached in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. At 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the rate of steam release required to remove decay heat plus reactor coolant pump heat (no LOOP case) is approximately 32 Ibm/sec. For the LOOP case (no reactor coolant pump heat) the steam release requirement is reduced to about 21.5 Ibm/sec.
For RCS conditions at 350 OF, the secondary steam pressure will be about 135 psia. At this pressure, it is estimated that a single steam generator PORV (atmospheric dump valve) can pass somewhere between 13-17 Ibm/sec, depending on the model used.
This means that both steam generator PORVs will be needed at the end of the cooldown. Assuming an equal distribution, at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> each PORV will be passing around 10-11 Ibm/sec for the LOOP case.
For a 10" discharge pipe and assuming isoenthalpic expansion of 10 Ibm/sec from saturated steam at 135 psi to superheated steam at atmospheric conditions, the steam exit velocity is estimated at about 560 ft/sec or 170 meters/sec. 170 meters per second is well above the required threshold of 28.6 meters per second to qualify for the factor of five reduction in the PORV atmospheric dispersion factors.
For a locked rotor accident (LRA), the heat in the reactor core which has to be removed through cooldown with the steam generator PORVs is the same as the heat released during a SGTR accident. The primary difference is that there are 3 steam generator PORVs available for cooldown during a LRA instead of the 2 flowing steam during a SGTR accident. This means that the PORV steam flow during a LRA cooldown will be 2/3 of the steam flow during a SGTR accident cooldown. Because the vertical velocity of the SGTR PORV steam flow is so high, even 2/3 of this will be far higher than the threshold required to justify a reduction of the atmospheric dispersion factors by a factor of 5. Therefore, the PORV atmospheric dispersion factors used to model the control room dose during a LRA can also be reduced by a factor of 5.
Conclusion At the end of the steam generator tube rupture accident, the vertical velocity of the steam exiting the unaffected steam generator PORVs exceeds the threshold required to qualify for the factor of five reduction in the PORV atmospheric dispersion factors.
Additionally, at the end of a locked rotor accident, the vertical velocity of the steam exhausting to atmosphere via the three steam generator PORVs also exceeds the 28.6 meters/sec requirement to qualify for the factor of five reduction in PORV atmospheric dispersion factors.
Letter No.04-494 Docket Nos. 50-338/339 Attachment Revised Input Data Pages 58 and 68 From Letter Dated September 12, 2003 (Serial No.03-464)
Letter No.04-494 Docket Nos. 50-338/339 Table 3.3-4: Analysis Assumptions and Key Parameter Values Employed in the SGTR Analysis Primary and Secondary Side Parameters Primary system volume (cubic feet)
SG Steam volume (cubic feet/SG)
SG Liquid volume (cubic feet/SG)
SG Liquid mass (gm/SG)
Control room volume (cubic feet)
Primary System Mass (lb or gm)
Secondary System Mass (lb or gm per generator)
Steam Mass Dilution 9786 3838 2054 4.43E+07 2.30E+05 4.37845E+05 lb or 1.986E+08 gm.
7200 lb/SG or 3.2666+06 gm1SG 2.81 E+05 Full Power Properties Temperature (degrees F)
Pressure (psia)
Density (gm/cc)
Steam Generator 525.24 850 0.76096 RCS Coolant Liquid 580.8 2250 0.71669 SGTR Flow Rates (all flow rates are in cubic feet per minute)
LOOP - Affected SG Time RCS to SG RCS to SG Steam SG Liquid to SG Steam to Liquid Steam Environment 0 - 103 sec 9.0961+01 1.164E+01 1.4946+03 0.000E+00 103 - 232 sec 7.6591+01 3.4706+00 1.972E+02 4.994E+03 232 -1800 sec 7.878E+01 4.860E+00 I
1.3086+02 3.3131+03 LOOP - Unaffected SG Time RCS to SG Liquid SG Liquid to Environment 0- 103 sec 0.1337 0
103 - 232 sec 0.1337 356.09 232 -1800 sec 0.1337 100.81 1800 - 2 hrs 0.1337 53.80 2 hrs - 8 hrs 0.1337 38.71 NO-LOOP - Affected SG Time RCS to SG RCS to SG Steam SG Liquid to SG Steam to Liquid Steam Environment 0 - 107 sec 8.021E+01 1.838E+01 1.494E+03 0.000E+00 107 - 196 sec 8.505E+01 3.190E+00 IX 2.726E+02 6.904E+03 196-1800 sec 8.228+01 2.460+00 1.521E+02 3.8506+03 NO-LOOP - Unaffected SG Time RCS to SG Liquid SG Liquid to Environment 0 - 107 sec 0.1337 0
107-196 sec 0.1337 484.85 196-1800sec 0.1337 184.30 1800 - 2 hrs 0.1337 65.49 2 hrs - 8 hrs 0.1337 50.35 Indicates revisions from original submittal I
Page 58
Letter No.04-494 Docket Nos. 50-338/339 Table 3.5-2: Analysis Assumptions & Key Parameter Values Employed In the Locked Rotor Analysis NSSS Parameters Core Power Number of Fuel Assemblies Primary System (RCS) Volume Steam Generator Liquid Volume Steam Generator Steam Volume Radial Peaking Factor Fuel Failure During Event RG-1.183 Non-LOCA Gap Fractions Main Control Room (MCR) Parameters 2958 MNVt 157 9786 ft3 6162 ft3 11,514 ft3 1.65 13%
See Table 3.5-1 2.30E5 ft3 3500 CFM Free Volume Normal Intake and Exhaust Flow Rate No Isolation of the Control Room Onsite Atmospheric Dispersion Factors Main Control Room Normal Intake 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 2.08E-3 seC/m 3 1.64E-3 sec/m3 6.46E-4 sec/m3 4.50E-4 sec/M3 3.36E-4 sec/m 3 Leak Rates Description Flow rate (cfm) 0.1337 Primary to Secondary Leakage SG liquid to steam 0 hr 0.25 hr 0.5 hr I hr 2 hr 8 hr 441.6 249.9 65.5 50.3 0
SG steam to Environment Ohr 0.25 hr 0.5 hr I hr 2 hr 8 hr 11183 6328 1658 1275 0
1 Indicates revisions from original submittal Page 68