ML061070039

Corrections to Safety Evaluation for Amendment No. 271 and No. 252, Respectively, Alternate Source Term

ML061070039
Person / Time
Site:	Salem
Issue date:	05/02/2006
From:	Stewart Bailey Plant Licensing Branch III-2
To:	Levis W Public Service Enterprise Group
Bailey S , NRR/ADRO/DORL, 415-1321
References
TAC MC3095, TAC MC3390
Download: ML061070039 (11)

Text

May 2, 2006 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, CORRECTIONS TO SAFETY EVALUATION FOR AMENDMENT NO. 271 AND NO. 252, RESPECTIVELY, RE: ALTERNATE SOURCE TERM (TAC NOS.

MC3390 AND MC3095)

Dear Mr. Levis:

The Nuclear Regulatory Commission (NRC) is issuing corrections to the Safety Evaluation (SE) for Amendment Nos. 271 and 252 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), respectively, in order to clarify certain statements in the SE. The amendment incorporated a full-scope application of an alternate source term methodology in accordance with Title 10 of the Code of Federal Regulations, Section 50.67 Accident Source Term. Amendment Nos. 271 and 252 were issued on February 17, 2006.

Following receipt of Amendment Nos. 271 and 252, your staff informed the NRC staff of several portions of the SE that required correction or clarification. In order to correctly reflect the current licensing basis for Salem, the NRC has revised the SE as follows:

(1) Page 5, 5th paragraph - The SE was clarified to state that the basis for reducing calculated atmospheric dispersion coefficients is that the time-dependent vertical velocity exceeds the 95th percentile wind speed by a factor of five, in accordance with the guidance in Regulatory Guide 1.194, ?Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants.

(2) Tables 1, 5, 6, 7, and 8 - The tables were updated to more accurately reflect the analyses assumptions and results.

Revised SE pages are attached. The changes are indicated by marginal lines. The NRC staff has determined that the corrections to the original SE do not change our previous conclusions regarding the acceptability of the changes approved in Amendment Nos. 271 and 252.

W. Levis If you have an questions, please contact me at (301) 415-1321 or at snb@nrc.gov.

Sincerely,

/RA/

Stewart N. Bailey, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosure:

Revised Safety Evaluation pages cc w/encl: See next page

Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:

Mr. Michael Gallagher Jeffrie J. Keenan, Esquire Vice President - Eng/Tech Support PSEG Nuclear - N21 PSEG Nuclear P.O. Box 236 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Lower Alloways Creek Township Mr. Dennis Winchester c/o Ms. Mary O. Henderson, Clerk Vice President - Nuclear Assessment Municipal Building, P.O. Box 157 PSEG Nuclear Hancocks Bridge, NJ 08038 P.O. Box 236 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs Mr. Thomas P. Joyce NJ Department of Environmental Site Vice President - Salem Protection and Energy PSEG Nuclear CN 415 P.O. Box 236 Trenton, NJ 08625-0415 Hancocks Bridge, NJ 08038 Mr. Brian Beam Mr. George H. Gellrich Board of Public Utilities Plant Support Manager 2 Gateway Center, Tenth Floor PSEG Nuclear Newark, NJ 07102 P.O. Box 236 Hancocks Bridge, NJ 08038 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. Carl J. Fricker 475 Allendale Road Plant Manager King of Prussia, PA 19406 PSEG Nuclear - N21 P.O. Box 236 Senior Resident Inspector Hancocks Bridge, NJ 08038 Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Mr. Darin Benyak Drawer 0509 Director - Regulatory Assurance Hancocks Bridge, NJ 08038 PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038

W. Levis If you have an questions, please contact me at (301) 415-1321 or at snb@nrc.gov.

Sincerely,

/RA/

Stewart N. Bailey, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosure:

Revised Safety Evaluation pages cc w/encl: See next page DISTRIBUTION:

RidsNrrDorlLpl1-2 RidsOgcRp RidsRgn1MailCenter Public RidsNrrPMSBailey RidsAcrsAcnwMailCenter LPL1-2 R/F RidsNrrLACRaynor WBeckner RidsNrrDorl GHill (2) Tech Branch Accession Number: ML061070039 OFFICE LPL1-2/GE LPL1-2/PM LPL1-2/LA AADB/BC LPL1-2/BC NAME CSanders/em SBailey CRaynor MKotzalas DRoberts (MHart for)

DATE 4/21/06 5/2/06 4/19/06 5/2/06 5/2/06 Official Record Copy

The licensee modeled two release scenarios, MSSV/ARV releases and PAPRP releases, and two receptor locations (one for each CR air intake). The licensee calculated /Q values for each release-receptor combination to identify the more favorable air intake (i.e., the air intake with the lower /Q value) for each scenario. The /Q values for the more favorable CR air intake were generally used to evaluate filtered air makeup during the pressurization mode, while the /Q values for the less favorable (i.e., higher /Q value) CR air intake were generally used to evaluate unfiltered air makeup prior to the pressurization mode and the unfiltered inleakage during the pressurization mode.

MSSV and ARV Releases Both Salem units are Westinghouse 4-loop pressurized water reactors with four steamlines for each unit. Each steamline has a set of five, self-actuated, spring-loaded MSSVs and one pneumatically-operated ARV. The MSSVs only open at or above their setpoint pressures, whereas the ARVs can be manually operated. Due to the clustering of the MSSVs and ARVs within each set, the licensee represented each set as one release point (the closest MSSV).

Also, based on the proximity of the MSSV and ARV sets to the CR air intakes, the licensee modeled the two closest MSSV/ARV sets.

To model the MSSV/ARV releases using ARCON96, the licensee assumed the MSSV/ARV releases were ground-level point sources. This is conservative because the MSSV and ARV releases are energetic releases from uncapped and vertically oriented vent pipes, so that additional plume rise will occur due to the buoyancy and momentum effects. The licensee determined that the exit velocity from the MSSV with the lowest pressure setpoint is sonic at 448 meters per sec (1002 miles per hour). Nonetheless, the licensee based the MSSV/ARV release height on the height of the top of the MSSV/ARV vent pipes and the source-to-receptor distance on the closest MSSV vent pipe horizontal distance to the CR intake.

The closest release-receptor combination is the Salem Unit No. 1 MSSV to the Salem Unit No. 1 CR intake. The MSSV release point is approximately 6.38 meters (20.9 feet) horizontally and 5.09 meters (16.7 feet) vertically above the CR intake, for a total distance of approximately 8.16 meters (26.8 feet). RG 1.194 states that any release-receptor combination closer than approximately 10 meters should be addressed on a case-by-case basis. However, due to the conservatism in ignoring the plume rise effects, the staff finds the licensees modeling to be acceptable.

RG 1.194 allows the ground level /Q values calculated with ARCON96 (on the basis of the physical height of the release point) to be reduced by a factor of 5 if (1) the release point is uncapped and vertically oriented, and (2) the time-dependent vertical velocity exceeds the 95th percentile wind speed by a factor of 5. The sonic exit velocity from the MSSV will be l considerably higher than the 10-meter (33-foot) 95th percentile wind speed value. Therefore, l the licensee reduced the resulting ARCON96 MSSV /Q values by a factor of five.

In its RAI dated May 5, 2005, the NRC staff asked whether a stuck-open MSSV or ARV is part of Salems licensing basis. By letter dated June 13, 2005, the licensee responded that Salems current licensing basis does not consider stuck-open MSSVs or ARVs. Also, for certain accident scenarios, the operators can cool down the plant by releasing steam from the ARVs.

As the transient progresses and secondary side pressure drops, the ARV plume will have a Enclosure

Table 1 Radiological Consequences Expressed as TEDE (1)

(rem)

Design Basis Accidents EAB (2) LPZ (3) Control Room (CR)

Loss-of-Coolant Accident 4.08 1.35 4.17 Dose criteria(4) 25 25 5.0 Main steamline break accident (4) 8.68E-2 2.88E-2 1.27E-1 Dose criteria 2.5 2.5 5.0 Main steamline break accident (5) 5.28E-1 1.87E-1 8.93E-1E-1 Dose criteria 25 25 5.0 Steam generator tube rupture (4) 2.21 3.31E-1 5.18E-1 l Dose criteria 2.5 2.5 5.0 Steam generator tube rupture (5) 1.57 3.29E-1 5.85E-1 l Dose criteria 25 25 5.0 Locked rotor accident 1.26 5.08E-1 1.30 l Dose criteria 2.5 2.5 5.0 Rod ejection accident 2.53E-1 1.31E-1 1.36 l Dose criteria 6.3 6.3 5.0 Waste Gas Decay Tank Rupture 4.10E-2 5.86E-3 2.07E-2 (6) l Dose criteria 2.5 2.5 5.0 (1)

Total effective dose equivalent (2)

Exclusion area boundary (3)

Low population zone (4)

Accident initiated iodine spike (5)

Pre-accident iodine spike (6)

No credit for CR isolation

Table 5 Parameters and Assumptions Used in Radiological Consequence Calculations for Main Steamline Break Accident Parameter Value Pre-incident iodine spike activity 60 µCi/gm dose equivalent I-131 Coincident spike appearance rate, based on Reactor coolant systems (RCS) letdown flow rate, gpm 165 RCS letdown demineralizer efficiency, % 90 l Coincident spike multiplier 500 Iodine spike duration, hrs 8 Total primary-to-secondary leakage l through all SGs, gpm 1 l Maximum primary-to-secondary leakage l through any one SG, gpd 500 l Duration, hours 32 Liquid Masses RCS 2.5E+8 gm Steam Generator (SG) (each) 1911 ft3 l Steam release from faulted SG, lbm 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 128,000 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0 Steam release from intact SGs, lbm 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5.00E+5 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.52E+5 8 to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> 2.01E+6 l Steam iodine partition coefficient in SGs Faulted SG (elemental and organic) 1.0 Unaffected SG 1.0 Elemental 1.0 Organic 1.0 Release points penetration area pressure relief panels and main steam l safety relief valves/atmospheric l relief valves l

Table 6 Parameters and Assumptions Used in Radiological Consequence Calculations for Steam Generator Tube Rupture Accident Parameter Value Pre-incident iodine spike activity 60 µCi/gm dose equivalent I-131 Coincident spike appearance rate, based on RCS letdown flow rate, gpm 165 RCS letdown demineralizer efficiency, % 90 l Coincident spike multiplier 335 Iodine spike duration, hrs 8 Total primary-to-secondary leakage l through all SGs, gpm 1 l Maximus primary-to-secondary leakage l through any one SG, gpd 500 l Duration, hours 32 Liquid Masses RCS 2.5E+8 gm SG (each) 1.19E+5 lbm (Unit 1) 1.28E+5 lbm (Unit 2)

Steam release from faulted SG, lbm 0 to 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 5.65E+4 lbm (flashed)

Steam release from intact SGs, lbm 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.65E+5 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.055E+6 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.50E+6 24 to 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> 4.77E+5 30 to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> 1.50E+5 Steam iodine partition coefficient in SGs Faulted SG 1.0 Unaffected SG 10 Release points main steam safety valves/ l atmospheric relief valves

Table 7 Parameters and Assumptions Used in Radiological Consequence Calculations for Locked Rotor Accident Parameter Value Fraction of failed fuel 0.05 Fraction of Core Inventory in Gap Kr-85 0.10 I-131 0.08 l Other noble gases / iodines 0.05 l Iodine speciation CNMT Secondary Aerosol 0.95 0 Elemental 0.0485 0.97 Organic 0.0015 0.03 l Total primary-to-secondary leakage l through all SGs, gpm 1 l Primary-to-secondary leakage duration, hours 32 Steam partition coefficient in SGs 10 l Steam release from all 4 SGs, lbm 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.55E+5 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 5.40E+5 8 to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> 2.40E+6 Release points main steam safety valves/ l atmospheric relief valves

Table 8 Parameters and Assumptions Used in Radiological Consequence Calculations for Control Rod Ejection Accident Parameter Value l

Fraction of rods that exceed DNB (melted fuel) 0.0025 l Gap fraction, all nuclide groups 0.10 Melt isotopic composition CNMT SG Noble gases 1.0 1.0 Iodine 0.25 0.5 Iodine species fraction CNMT SG Particulate/aerosol 0.95 0 Elemental 0.0485 0.97 Organic 0.0015 0.03 Containment free volume, ft3 2.62E+6 Containment Sprays Not credited Containment release 0-24 hours, %/day 0.1 24-720 hours, %/day 0.05 l

Duration of release, days 30 l Total primary-to-secondary leakage l through all SGs, gpm 1 l Primary-to-secondary leakage duration, hours 8 Steam generator mass (each) 1.19E+5 lbm (Unit 1) 1.28E+5 lbm (Unit 2)

Steam partition coefficient in SGs 1 Steam release from 4 SGs, lbm 0- 110 seconds 5.12E+5 Release point Containment leakage plant vent l Secondary main steam safety valves/ l atmospheric relief valves